ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 106

Document Title:

Summary Site Profile Document for the 

Brookhaven National Laboratory 

Document Number: ORAUT-TKBS-0048

Revision: 00

Effective Date: 08/30/2006

Type of Document: TBD

Supersedes: None 

 

Subject Experts: Lori Arent, Samuel Chu, John A. Devanney, Leo Faust, Robert L. Morris, 

and Eugene Potter

Document Owner

Approval: Signature on File Approval Date: 08/18/2006

Melton H. Chew, TBD Team Leader

Approval: Signature on File Approval Date: 08/11/2006

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 08/15/2006

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 08/17/2006

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 08/30/2006

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 2 of 106

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

08/30/2006 00 New Site Profile for the Brookhaven National Laboratory. First 

approved issue. Incorporates internal and NIOSH formal review 

comments. Training required: As determined by the Task Manager. 

Initiated by Melton H. Chew.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 3 of 106

TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations ...............................................................................................................

PAGE

7

1.0 Introduction ............................................................................................................................10

1.1 Purpose..........................................................................................................................11

1.2 Scope.............................................................................................................................11

1.2.1 Site Description ................................................................................................... 11

1.2.2 Occupational Medical Dose................................................................................. 11

1.2.3 Occupational Environmental Dose ...................................................................... 11

1.2.4 Occupational Internal Dose ................................................................................. 12

1.2.5 Occupational External Dosimetry ........................................................................ 12

2.0 Site Description......................................................................................................................13

2.1 Introduction ....................................................................................................................13

2.2 Purpose..........................................................................................................................13

2.3 Scope.............................................................................................................................13

2.4 Site Description and General Information.......................................................................13

2.4.1 Brookhaven Graphite Research Reactor, Building 701........................................ 16

2.4.2 High Flux Beam Reactor (HFBR), Building 750................................................... 20

2.4.3 Medical Research Center (MRC), Building 490 ................................................... 21

2.4.4 Brookhaven Medical Research Reactor, Building 491 ......................................... 21

2.4.5 Positron Emission Tomography, Building 906 ..................................................... 22

2.4.6 Cosmotron, Building 902 ..................................................................................... 22

2.4.7 Alternating Gradient Synchrotron, 900 Series Buildings ...................................... 22

2.4.8 Brookhaven LINAC Isotope Producer, Building 931 ............................................ 27

2.4.9 Tandem Van de Graaff, Building 901................................................................... 27

2.4.10 National Synchrotron Light Source, Building 725................................................. 29

2.4.11 Relativistic Heavy ion Collider Complex, Building 1000 Series ............................ 29

2.4.12 Hot Laboratory, Building 801 ............................................................................... 32

2.4.13 High Intensity Radiation Development Laboratory (HIRDL), Building 830............ 33

2.4.14 Chemistry Department, Building 555 ................................................................... 33

2.4.15 Physics Department, Building 510....................................................................... 34

2.4.16 Other Miscellaneous Radiological Research Facilities......................................... 35

 2.4.16.1 LMFR Support, Building 820 ............................................................... 35

 2.4.16.2 Hot Machine Shop, Building 530......................................................... 35

 2.4.16.3 Health Physics & Safety Instrumentation, Building 535 ....................... 35

2.4.17 Waste Management Facility (WMF), Radwaste Reclamation Building, 

 Building 865........................................................................................................ 35

2.4.18 Radiological Waste Decontamination Facility, Building 650................................. 36

2.5 Isotopes of Concern .......................................................................................................37

 2.6 Magnitude of Site Activity ...............................................................................................41

 2.7 Major Incidents............................................................................................................. 46 

 2.8 Radiological Access Controls....................................................................................... 46 

References ........................................................................................................................................47

General Reference: ...........................................................................................................................48
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Crease, Robert P., Making Physics, A Biography of Brookhaven National Laboratory, 1946 -1972, 

University of Chicago Press, ISBN 0 226 12017, 1999 ...................................................................... 48

3.0 Occupational Medical Dose.................................................................................................... 49

3.1 Introduction ....................................................................................................................49

3.2 Purpose..........................................................................................................................49

3.3 Scope.............................................................................................................................49

3.4 Examination Frequency..................................................................................................49

3.5 Equipment and Techniques............................................................................................50

3.6 Entrance skin exposure (ESE) .......................................................................................51

3.7 Organ Dose Calculations ...............................................................................................52

 3.7.1 Organ Dose from Collimated Exposure ............................................................... 53

 3.7.2 Organ Dose from Uncollimated Exposure ........................................................... 53

 3.7.3 Skin dose ............................................................................................................ 53

3.8 Dose Reconstruction......................................................................................................55

3.9 Uncertainty.....................................................................................................................55

References ........................................................................................................................................61

4.0 Occupational Environmental Dose ......................................................................................... 62

4.1 Introduction ....................................................................................................................62

4.2 Purpose..........................................................................................................................62

4.3 Scope.............................................................................................................................62

4.4 Ambient External Radiation............................................................................................62

 4.4.1 Use of Off-Site Monitoring ................................................................................... 62

 4.4.2 Applicability For Dose Reconstruction ................................................................. 63

4.5 Inhalation of Onsite Airborne Radionuclides...................................................................63

 4.5.1 Stack Monitoring ................................................................................................. 63

4.6 Ingestion ........................................................................................................................68

4.7 Uncertainty.....................................................................................................................68

References ........................................................................................................................................69

5.0 Occupational Internal Dose.................................................................................................... 72

5.1 Introduction ....................................................................................................................72

5.2 Purpose..........................................................................................................................72

5.3 Scope.............................................................................................................................72

5.4 The bioassay records in the individual files ....................................................................72

 5.4.1 Brookhaven National Laboratory Bioassay Record (Form 1720) ......................... 72

 5.4.2 Brookhaven National Laboratory Tritium Exposure Evaluation ............................ 73

 5.4.3 Summary of Whole-body Counting Results ......................................................... 73

 5.4.4 Individual Whole-body Counting Results ............................................................. 73

5.5 In Vitro Minimum Detectable Activities, Counting Methods, and Reporting Protocols ..... 74

 5.5.1 In Vitro Urine Analysis ......................................................................................... 74

 5.5.2 History of In Vitro Urine Analysis ......................................................................... 74

 5.5.3 Fecal Sample Analysis ........................................................................................ 75

5.6 In Vivo Minimum Detectable Activities, Counting Methods, and Reporting Practices...... 75

 5.6.1 Whole-Body Counting ......................................................................................... 75

 5.6.2 Chest Counting ................................................................................................... 77

 5.7 Uncertainty.....................................................................................................................77

5.8 Detection Limits..............................................................................................................77

 5.9 Excreta Sample Kit Codes................................................................................................80
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5.10 Solubility Type, Fraction Activity, and Particle Size by Facility....................................... 81

5.11 Facility-Specific Radionuclide Conversions ...................................................................84

5.12 Workplace Monitoring Data ...........................................................................................84

5.13 Source Term Data.........................................................................................................84

5.14 Radon ...........................................................................................................................84

References ........................................................................................................................................86

6.0 Occupational External Dosimetry ........................................................................................... 91

6.1 Introduction ....................................................................................................................91

6.2 Purpose..........................................................................................................................91

6.3 Scope.............................................................................................................................91

6.4 BNL Dosimetry Program ................................................................................................91

 6.4.1 Neutron Dosimetry .............................................................................................. 91

 6.4.2 Special Dosimetry ............................................................................................... 91

 6.4.3 Lost or Destroyed Dosimeters ............................................................................. 92

6.5 Missed Dose ..................................................................................................................92

 6.5.1 Unmonitored Worker ........................................................................................... 92

6.6 Radiation Energies And Percentages At Selected BNL Facilities ................................... 93

6.7 Neutron Radiations And Percentages ............................................................................94

6.8 Recorded Dose Practices...............................................................................................94

6.9 Interpretation of Reported Doses....................................................................................95

 6.10 Adjustments to Recorded Dose........................................................................................ 95

6.11 Bias and Uncertainty........................................................................................................ 95

References ........................................................................................................................................98

Glossary .......................................................................................................................................... 100
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LIST OF TABLES

TABLE TITLE

2-1 Site information of major nuclear research facilities at BNL....................................................

PAGE

18

2-2 Area information and parameters of nuclear operations ......................................................... 37

2-3 Magnitude of operation .......................................................................................................... 41

2-4 Major incidents.......................................................................................................................46

3-1 Parameters and organ doses for 14-in. by 17-in. PA chest radiography................................. 56

3-2 Parameters and organ doses for 14-in. by 17-in. LAT chest radiography ............................... 58

3-3 Maximum organ doses for 14-in. by 17-in. PA chest radiography ........................................... 60

3-4 Maximum organ doses for LAT chest radiography ................................................................. 60

4-1 External gamma radiation dose.............................................................................................. 63

4-2 Annual environmental occupational radionuclide inhalation.................................................... 65

5-1 Internal dose control program (in vitro) ................................................................................... 75

5-2 Internal dose control program (in vivo) ................................................................................... 77

5-3 Detection limits.......................................................................................................................78

5-4 Excreta sample kit codes ....................................................................................................... 80

5-5 Solubility type, fraction activity, and particle size by facility..................................................... 81

5-6 Facility-specific radionuclide conversions............................................................................... 84

6-1 Estimated annual missed photon, beta, and neutron dose ..................................................... 92

6-2 Selection of beta (electron) and photon energies and percentages ........................................ 93

6-2a Beta and photon energies and percentages (support facilities) .............................................. 93

6-3 Facilities, neutron energies, percentages, and correction factors ........................................... 94

6-4 Recorded dose practices........................................................................................................ 94

6-5 Interpretation of reported data................................................................................................ 95

6-6 Adjustments to recorded dose................................................................................................ 95

6-7 Bias and uncertainty...............................................................................................................97

LIST OF FIGURES

FIGURE TITLE

2-1 Aerial view of Brookhaven National Laboratory ......................................................................

PAGE

14

2-2 Nuclear and radioactivity sites at BNL.................................................................................... 17

2-3 The AGS Complex ................................................................................................................. 23

2-4 The Relativistic Heavy Ion Collider (RHIC) Complex .............................................................. 30

2-5 RHIC ring intersection points.................................................................................................. 31
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ACRONYMS AND ABBREVIATIONS

ADD Average Depth Dose

AEC U.S. Atomic Energy Commission

AF Absorption Factor

AGS Alternating Gradient Synchrotron

AMAD activity median aerodynamic diameter

AP anterior-posterior

ATR AGS-To-RHIC transfer line

BF Backscatter factor

BGRR Brookhaven Graphite Research Reactor

BIRC Brookhaven Isotope Research Center

BLAF Brookhaven Lab Animal Facility

BLIP Brookhaven Linac Isotope Producer

BMRR Brookhaven Medical Research Reactor

BNCT Boron Neutron Capture Therapy

BNL Brookhaven National Laboratory

Bq Becquerel

CBA Colliding Beam Accelerator

cGy centigray

CLIF Chemistry Linac Irradiation Facility (also CLIP in some documents)

cm centimeter

CNF Cold Neutron Facility

CR-39 Columbia resin, number -39

D&D Decontamination and Decommissioning

DAS Department of Applied Science

DAT Department of Applied Technology

DCF dose conversion factor

DE Dose Equivalent

DOE U.S. Department of Energy

DOELAP Department of Energy Laboratory Accreditation Program

Dpm disintegration per minute

DR dose rate

EEOICPA Energy Employees Occupational Illness Compensation Program Act

EML Environmental Measurement Laboratory

ENSD Entrance Skin Dose

ENSDNPB Entrance Skin Dose Near Primary Beam

ESE entrance skin exposure

EXSD Exit Skin Dose

EXSDNPB Exit Skin Dose Near Primary Beam

F Fast (solubility rate)

FDA Food and Drug Administration

G Gram

GIF Gamma Irradiation Facility

Gy gray
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HASL Health and Safety Laboratory

HEBT High Energy Beam Transport

HFBR High Flux Beam Reactor

HIRDL High Intensity Radiation Development Laboratory

HITL Heavy Ion Transfer Line

HPI Health Physics Instruction

hr hour

HVL Half Value Layer

ICRP International Commission on Radiological Protection

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt (1,000 electron-volts)

kVp kilovolts-peak

L liter

LAF Laser Electron Accelerator Facility

LAT lateral 

LEAF Low Energy Accelerator Facility

LINAC linear accelerator

LMFR Liquid Metal Fuel Reactor

M moderate (solubility rate)

mA milliampere

MDA minimum detectable activity, or amount

MDL minimum detectable level

MEL Metallurgical Evaluation Laboratory

MeV million electron volt

MFP mixed fission products

mg milligram

mGy milligray

min minute

mL milliliter

mm millimeter

mR milliroentgen

mrad milliard

MRC Medical Research Center

mrem millirem

ms millisecond

MW megawatts

NaI (TL) Sodium-iodide, thallium-activated scintillation detector

NBTF Neutral Beam Test Facility

nCi nanocurie

NCRP National Council on Radiation Protection

NIOSH National Institute for Occupational Safety and Health

NSLS National Synchrotron Light Source

NSRL NASA Space Radiation Laboratory

NTA nuclear track emulsion, type A

NVLAP National Voluntary Laboratory Accreditation Program

NYOO New York Operations Office of the AEC
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ORAU Oak Ridge Associated Universities

OW open window

PA posterior-anterior

pCi picocurie

PET Positron Emission Tomography

QF quality factor

R roentgen

RARAF Radiological Research Accelerator Facility

RBE relative biological effectiveness

REF Radiation Effects Facility

RHIC Relativistic Heavy Ion Collider

RSD Remote Skin Dose

RTF Radiation Therapy Facility

S slow (solubility rate)

SD standard deviation

sec second

SF spontaneous fission

SID source-to-image distance

SSD source-to-skin distance

ST Severn-Trent

TBD technical basis document

TPL Target Processing Laboratory

U.S.C. United States Code

VUV Vacuum Ultraviolet

WBC whole-body counter

WBNIF Whole Body Neutron Irradiation Facility

WMF Waste Management Facility

yr year

μCi microcurie

μg microgram
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1.0

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist the NIOSH staff in the completion of the individual work required for each dose 

reconstruction.

INTRODUCTION

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 

U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. 

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally derived radiation

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

reconstruction. NIOSH, however, does not consider the following exposures to be occupationally 

derived:

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees

with a covered cancer or restrict the “duty” to nuclear weapons work.

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
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• Radiation from naturally occurring radon present in conventional structures

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

1.1 PURPOSE

The purpose of this document is to provide a site profile that contains technical basis information for 

the evaluation of the total occupational dose for EEOICPA claimants who were employed at the 

Brookhaven National Laboratory.

1.2 SCOPE

This site profile consists of six sections: Introduction, Site Description, Occupational Medical Dose, 

Occupational Environmental Dose, Occupational Internal Dose, and Occupational External Dosimetry.

1.2.1

Section 2, Site Description, of this site profile describes the features and the history of the Brookhaven 

National Laboratory (BNL) including site areas or buildings, site processes, periods of operation, 

radionuclides of concern and other information pertinent to dose reconstruction. 

Site Description

This section provides a description and general information for the various facilities that operated, or 

are operating at the BNL. These include the Brookhaven Graphite Research Reactor and its 

replacement, the High Flux Beam Reactor; the operations of the Medical Research Center, the 

positron emission tomography facility and the medical research reactor; the operation of the various 

accelerators and the production of radiopharmaceuticals. The section also addresses some of the 

high-energy physics facilities such as the relativistic heavy ion collider complex where ion beams 

traveling in opposite directions in a circular orbit collide.

1.2.2

Section 3, Occupational Medical Dose, of this site profile provides information regarding the historical 

BNL occupational x-ray program. Much of this information was obtained through personal interviews 

with the clinicians and x-ray technicians. Beginning in 1947 and continuing through at least 1964, 

reactor operators and possibly hospital medical staff were subjected to an entry and annual physical 

examination including a PA chest x-ray. All other workers, including visiting scientists employed for 

more than three months received an entry physical and re-examination, including a PA chest x-ray on 

a frequency averaging about one exam every 2.5 years. The section presents organ doses for both 

PA and LAT chest films using collimated and uncollimated exposures. Skin doses both inside and 

outside the primary beam are determined for PA and LAT chest projections.

Occupational Medical Dose

1.2.3

Section 4, Occupational Environmental Dose, discusses the radiation doses received by workers at 

the BNL but outside the facilities. From 1967 to 1984 ambient gamma radiation dose rates are 

available from four perimeter stations surrounding the site at NW, SW, SE, and NE coordinates. 

Beginning in 1985 the onsite monitoring program was expanded to approximately 20 onsite stations. 

The ambient external gamma radiation outside the radiological controlled areas is the result of 

background radiation or gamma emissions from stack effluent such as Ar-41 or skyshine due to air 

scatter from an otherwise well-shielded radiation source. The radionuclides emitted from the BNL 

stacks are identified and their annual inhalation (Bq/yr) listed. Neither potable groundwater nor soil 

ingestion have been found to be pathways of exposure. No ingestion dose is indicated.

Occupational Environmental Dose
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1.2.4

Section 5, Occupational Internal Dose, discusses the internal dosimetry systems and practices 

employed at the BNL. The section provides supporting technical data to evaluate internal 

occupational doses that can reasonably be associated with worker radiation exposures covered by 

the EEOICPA legislation. These doses include occupational internal exposures in BNL facilities and 

onsite exposures to BNL environmental releases. It lists the types of routine monitoring that constitute 

the in vitro and the in vivo dose control programs as well as the minimum detectable activities, 

counting methods, and reporting practices. The section addresses the uncertainty for BNL exposure 

and dose records.

Occupational Internal Dose

1.2.5

Section 6, Occupational External Dosimetry, discusses the external dosimetry systems and practices 

at the BNL. It presents historical and current practices in relation to the evaluation of external 

exposure data for monitored and unmonitored workers. It contains supporting documentation to 

assist in the evaluation of occupational external exposures from processes that occurred at the BNL. 

In addition, it addresses the evaluation of worker exposure, missed dose, and the bias and uncertainty 

associated with the monitoring of external dose. The missed doses were basically those resulting 

from dosimeter minimum detectable levels and exchange periods, as well as the inability to determine 

true doses from high energy particles generated by the various accelerators. Neutron doses were 

measured with NTA film. Spallation products, that produced stars on the NTA film, were related to an 

effective neutron dose that was added to the total neutron dose.

Occupational External Dosimetry
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2.0

2.1 INTRODUCTION

SITE DESCRIPTION

BNL was founded in 1947, and it has been in operation since then at Upton, Long Island, New York, 

under a contract between the U.S. Atomic energy Commission (AEC) (which has now been assumed 

by the Department of Energy) and Associated Universities Inc. (AUI). The site was formerly Camp 

Upton and it was used by the Army during World Wars I and II.

BNL’s early research focused on advanced physics, but expanded into its current suite of research in 

the fields of medicine, biology, chemistry, physics, materials science, nuclear engineering, and 

environmental research. The BNL was established to provide facilities for scientific research and was 

organized into Departments that provide research nuclear reactors, various particle accelerators, and 

engineering facilities, and support the disciplines of Biology, Chemistry, Physics, Medicine, Applied 

Science, Accelerator and Applied Mathematics Departments. 

2.2 PURPOSE

The purpose of this section is to describe the facilities of the Brookhaven National Laboratory (BNL), 

the time and period of the process operations, and major site incidents that resulted in a significant 

internal or external exposure to one or more persons.

2.3 SCOPE

This section describes the features and history of the BNL including the site areas or buildings, site 

processes, periods of operation, radionuclides of concern, and other information pertinent to dose 

reconstruction.

2.4 SITE DESCRIPTION AND GENERAL INFORMATION

Figure 2-1 shows an aerial view of BNL. 
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Figure 2-1. Aerial view of Brookhaven National Laboratory. 

Research Reactors

The Laboratory’s scientific history began in 1950 with the operation of the Brookhaven Graphite 

Research Reactor (BGRR), a research reactor used for peaceful scientific exploration in the fields of 

medicine, biology, chemistry, physics, and nuclear engineering. The BGRR operated until 1969. In 

1965 its capacity was replaced and surpassed by the High Flux Beam Reactor (HFBR), which provide 

neutrons to researchers of all disciplines, from solid state physics to art history. 

Medical Research Center 

Medical research at BNL began in 1950 with the opening of one of the first hospitals devoted to 

nuclear medicine. It was followed by the Medical Research Center (MRC) in 1958 and the 

Brookhaven Medical Research Reactor (BMRR) in 1959. The BMRR is the first nuclear reactor built 

exclusively for medical and biological research and came on-line on March 15, 1959 and operated 

until October 2000. 

The Radiation Therapy Facility (RTF), operated jointly by the BNL Medical Department and State 

University of New York at Stony Brook, is a high-energy dual x-ray mode linear accelerator (LINAC) 

for radiation therapy of cancer patients. This accelerator was designed to deliver therapeutically 

beams of x-rays and electrons for conventional and advanced radiotherapy techniques.

Chemists and physicians teamed up to view the inner workings of the brain in 1977 with the advent of 

Positron Emission tomography (PET) cameras. Two more imaging techniques were added to the 
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PET research efforts to form the Center for Imaging and Neuroscience in 1996. Except for the 

BMRR, all of these facilities are all currently operating. 

Particle Accelerators

High energy particle physics research began in 1952 with the Cosmotron, the first particle physics 

accelerator to achieve billion-electro-volt energies. The Cosmotron operated from 1953 to 1966. In 

1960, the Alternating Gradient Synchrotron (AGS), a large accelerator, was built to surpass the 

Cosmotron’s capabilities. The AGS is capable of accelerating protons to energies up to 30 GeV and 

heavy ion beams to 15 GeV/amu. The AGS achieved full energy in 1960 and is still in use. 

Between 1967 and 1970, the Tandem Van De Graaff, 60-inch Cyclotron, and Vertical Accelerator

were used for medium energy physics investigations and for special isotope production. The heavy 

ions from the Tandem Van De Graaff can also be injected into the AGS for physics experiments. The 

Tandem Van De Graaff began operating in 1970 and continues operating to the present.

The Heavy Ion Transfer tunnel connects the coupled Tandem Van de Graaff and the AGS. The 

interconnection of these two facilities permits intermediate mass ions to be injected into the AGS 

where they can be accelerated to energy of 15 GeV/amu. These ions then are extracted and sent to 

the AGS experimental area for physics research. The AGS Booster is a circular accelerator with a 

circumference of 200 meters that receives either a proton beam from Linac or heavy ions from the 

Tandem Van De Graaff. The Booster accelerates proton particles and heavy ions before injecting 

them into the AGS ring. The Booster receives protons and heavy ions from the LINAC and Tandem 

Van De Graaff Facilities to increase their intensity for delivery to the AGS.

The Brookhaven Linac Isotope Producer (BLIP) became operational in 1973. Protons from the LINAC 

are sent via an underground beam tunnel to the BLIP facility where they strike various target metals. 

These metals, which become activated by the proton beam, are then processed at Target Processing 

Laboratory for use in radiopharmaceutical development and production. The targets are cooled by a 

continuously recirculating water system. The BLIP facility underwent significant upgrades between 

1994 in support of the Brookhaven Isotope Research Center (BIRC) program. The 200 MeV Proton 

Linear Accelerator serves as a proton injector for the AGS and also supplies a continuous beam of 

protons for radionuclide production by spallation reactions in the BLIP.

In 1982, the National Synchrotron light Source (NSLS) began operation. The NSLS guides charged 

particles in an orbit. As the electrons spin inside a hollow donut-shaped tube called an electron 

storage ring, they give off light called synchrotron light. This light, which can be detected by 

specialized instruments, is used to study the properties of matter. The NSLS utilizes a linear 

accelerator and booster synchrotron as an injection system for two electron storage rings which 

operate at energies of 750 MeV vacuum ultraviolet (VUV), and 2.5 GeV (x-ray). The synchrotron 

radiation produced by their stored electrons is used for VUV spectroscopy and for x-ray diffraction 

studies.

Brookhaven’s newest accelerator facility is the Relativistic Heavy ion Collider (RHIC), which was 

completed in 1999. The RHIC is designed to recreate a state of matter that scientists believe existed 

moments after the universe was formed. 

Department of Applied Science 

The Target Processing Laboratory (also call Hot Laboratory in 1993) officially opened January 15, 

1951. The original purpose of the central facility was to provide appropriately shielded areas for 

research with large amounts of radioactive material. The “hot” area of the Hot Laboratory included
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five hot cells, three chemical-processing hot cells and three high-level hot cells for handling and 

processing of radioactivity in gaseous, liquid, or solid form.

In 1971-1972, the High Intensity Radiation Development Laboratory (HIRDL), which contains millioncurie range Co-60 and Cs-137 sources, is used for source development and experimental process 

irradiations.

Waste Management

The new Waste Management Facility (WMF) replaces the original Hazardous Waste Management 

Facility in its entirety and consolidates several waste management operations into functional 

buildings. The WMF also provides for significant expansion within a dedicated site suitable for 

handling and storing hazardous and radioactive wastes generated at BNL. 

Figure 2-2 shows the location of the major sites where nuclear research activities were carried out at 

BNL.

Table 2-1 lists the major nuclear research facilities and their associated summary information.

The following subsections comprise further descriptions of the research facilities of significance on the 

BNL Site and are intended to provide background information relevant to the dose reconstruction 

efforts.

2.4.1

The operational BGRR and its support systems were housed in Buildings 701, 704, 708 and 709. The 

BGRR operated until 1969. At the present time, the BGRR facility is used as a visitor's center, 

museum, and office area for other projects.

Brookhaven Graphite Research Reactor, Building 701

Building 701 – 

The BGRR was a graphite moderated and reflected, thermal neutron, air cooled research facility. The 

original fuel loading was natural uranium which was replaced in later years by highly enriched fuel 

elements. The 25 foot, 700-ton graphite cube was built in two halves separated by a narrow vertical 

gap running east-west. Filtered cooling air was drawn into this gap and flowed through north-south 

channels removing heat from the fuel elements and graphite and then flowed out of the reactor 

through underground concrete ducts, passing through filters, coolers and primary exhaust fans which 

discharged into a 100 meter high stack.

The BGRR achieved criticality on August 22, 1950 and operated at power levels up to 28 MW with 

natural uranium fuel. In April of 1958, the reactor was reloaded with highly enriched uranium fuel 

elements and operated at power levels of up to 20 MW. On June 10, 1968, the use of the reactor for 

experimental purposes was terminated and initial decommissioning operations began. The last fuel 

shipped to the Savannah River processing plant in June 1972 and the canal was pumped dry. 

Building 709 – 

Building 709 is the fuel transfer and storage canal of the BGRR. During the 1950's, prior to the 

replacement of the natural uranium fuel slugs, there was a substantial problem with corrosion of fuel 

elements in the canal. This fuel was subject to deterioration and oxidation while in storage, making 

the slugs “dusty” when they were eventually prepared for shipment. The new type of fuel elements, 

which replaced the natural uranium elements in 1958, eliminated this problem. The new elements 

were aluminum-clad enriched uranium.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 17 of 106

Hot 

Laboratory

RHIC

Linac

Booster Accelerator

Alternating

Gradient

Synchrotron

Tandem

Van de Graaff

Medical 

Research 

Center 

(BMRR)

BGRR

HFBR

Chemistry 

Department

BLIP

Cosmotron

PET

NSLS

HIRDL

Physics 

Department

Figure 2-2. Nuclear and radioactivity sites at BNL. 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 18 of 106

Table 2-1. Site information of major nuclear research facilities at BNL. 

Section Name Description Period 

Building 

No. Note 

2.4.1 Brookhaven Graphite 

Research Reactor (BGRR)

Uranium fueled Graphite moderated 

and reflected Reactor

28 MW

1950-1969 701

2.4.1 BGRR fuel canal BGRR fuel storage 1950-1972 709

2.4.1 BGRR exhaust fan house Highly contaminated ducting and fans 1950-1972 704

BGRR exhaust filter house Contaminated HEPA filter elements 1950-

unknown

708 filter elements are still in 

place in 1997

2.4.2 High Flux Beam Reactor 

(HFBR)

30-60 MW thermal heavy water 

moderated research reactor

1965-1999 750 Building cease operation due 

to leak in spent fuel pool in 

1997

2.4.2 HFBR Stack 100 m stack and houses iodine filters 1965-1999 705

2.4.2 Cold Neutron Facility

(CNF)

Produce extremely low energy 

neutron by moderate neutron with 

liquid hydrogen

1980-

unknown

751

2.4.3 Medical Research Center 

(MRC)

Medical Department 1959-present 490 Dispersible radioisotopes 

PuBe sources

2.4.4 Brookhaven Medical 

Research Reactor (BMRR)

Enriched U fueled, light water 

moderated and cooled reactor

1959-2000 491

2.4.5 PET Imaging Laboratory Positron Emission Tomographic 

Scanners for nuclear medicine and 

neurosciences researches

1977- present 906 Use of positron emitters 

(Carbon-11, Fluorine-18, 

nitrogen-13 and oxygen-18)

2.4.6 Cosmotron Proton accelerator 1949-1968 902

2.4.7 Alternating Gradient 

Synchrotron (AGS) 

A large accelerator capable of 

accelerating protons to energies up to 

30 GeV and heavy ion beams to 15 

GeV/amu

1960-present 913 AGS tunnel, short-lived 

activation products in tunnel.

2.4.7 200 MeV LINAC Pre-accelerator and linear accelerator 

providing proton source

1967-present 930

2.4.7 LINAC cooling support 

service

Provides water-cooling services for 

the LINAC complex

1967-present 930 

MER

Contains activated water

2.4.7 AGS Booster Tunnel Accumulator booster ring 1960-present 942 short-lived activation 

products in tunnel

2.4.7 AGS Experiment Area Experimental target area for the AGS 

ring and later contains a proton beam 

switchyard system

1960-present 912 Activated components

2.4.7 AGS ventilation system fan 

houses

Ventilation exhaust fans 1960-present 913A-E High radiation area due to 

proximity to the tunnel. 

Activated water

2.4.7 AGS LINAC Contains 50 MeV AGS LINAC 1960s 914

2.4.7 E10 Power Supply Building Houses an assortment of power 

supply

1971-present 920 Radiation hazard from 

activated cooling water and 

oil in transformer.

2.4.7 Siemens motor generator 

house

Electrical equipment and lubricating 

equipment for the motor generator

1960-present 928,929 Radiation hazard from 

activated cooling water

2.4.7 AGS Department Office, field testing shops, machine 

shops, high bays.

1960-present 911 Stored and repair activated 

components.

2.4.7 g-2/Bubble Chamber Contains the 80” Bubble Chamber 1962-1978 919 A release of tritium occurred 

in 1973

2.4.7 Proton Target Proton target that produces 

secondaries for the g-2 ring

1962-present 949 Highly activated target

2.4.7 7’ Bubble Chamber High energy physics experimental 

facility

1972-1979 960 Activated components

2.4.7 AGS Warehouses Warehouses 1960-present 975,918,

196,209,

424,178

Stored activated 

components.

2.4.8 Brookhaven LINAC Isotope 

Producer (BLIP)

Production of radioisotopes via

irradiation in the LINAC beam (200 

Mev Protons)

1961-1962 931B Bldgs 931A and C have no 

nuclear hazard

2.4.9 Tandem Van de Graaff Van de Graaff accelerators

60” proton Cyclotron producing 

10 MeV protons, 20 MeV deuterons 

and 40 MeV alpha particles for 

medical radionuclide research

1970 -

present

901 Dynamitron (positron 

accelerator) 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 19 of 106

Table 2-1 (Continued). Site information of major nuclear research facilities at BNL.

Section Name Description Period 

Building 

No. Note 

2.4.10 National Synchrotron Light

Source (NSLS)

Produce intense sources of x-rays, 

ultraviolet and infra-red radiation

1981-present 725

2.4.10 NSLS Development 

Laboratory

Houses a linear accelerator 1981-present 729

2.4.11 Relativistic Heavy Ion 

Collider (RHIC) Complex

High-energy particle accelerator and 

collider

BRAHMS experimental hall

RHIC service building

Open area (future experimental area)

RHIC support building

Beam tunnel

Offices and tech shops

STAR Detector

Service building

Beam tunnel

PHENIX experiment

Experimental support building

PHOBOS experiment

Future experimental hall

1999-present

1981-present

1981-present

1996-present

1994-present

1981-present

1981-present

1985-present

1981-present

1981-present

1988-present

1985-present

1994-present

1994-present

1002

1004A

1004

1004B

1005

1005S

1006

1006A

1007

1008

1008A

1010

1012

Has prior radioactive works

Uranium plate machining

Has prior radioactive works

2.4.12 Hot Laboratory Target Processing Laboratory 1951-present 801 There was an explosion of 

UF6/BrF3 on 5/15/1957. Few 

people injured.

2.4.13 High Intensity Radiation 

Development Laboratory 

(HIRDL)

Radioactive sources fabrication, 

irradiation cells and since 1970, the 

Gamma Irradiation Facility (GIF) 

1951-present

GIF 1970-

present

830 The GIF contains 

approximately 35,000 Ci of 

Co-60 and million Ci Co-60 

and Cs-173 source

2.4.14 Chemistry Department Studies from chemical kinetics, to hotatom effects, to nuclear chemical 

studies of nuclear decay schemes 

and nuclear reaction cross-sections, 

to solar neutrino studies and 

radiocarbon dating.

1966-present 555

2.4.15 Physics Department laboratories, office space, a machine 

shop and basement area

1962-present 510 Radioactive material 

storage, counting labs 

2.4.16.1 LMFR Support Facility Research and development work for 

the Liquid Metal Fuel Reactor

1957-1975 820

2.4.16.2 Hot Machine Shop Hot Machine Shop 1947-1975 530

2.4.16.3 Instrument Division Research and development of x-ray 

and neutron detectors and 

corresponding electronics, which are 

used at the NSLS, the HFBR, the 

AGS, and other worldwide scientific 

research facilities.

1964-present 535

2.4.17 Radwaste Reclamation 

Building

Located within the Waste 

Management facility (WMF) for 

processing of radwaste

1997-present 865

2.4.17 Tritium Evaporator Facility Reduce tritium release by evaporation 1995-present 802B

2.4.17 Waste Management 

Incinerator

Low-level Waste incinerator 1995-present 444

2.4.18 Radiological Waste 

Decontamination Facility

Decontamination of radiological waste 1959-1996 650

During the early years of operation of the BGRR, fuel failures occurred which resulted in radioactive 

materials being released to the air stream that cooled the reactor. There were 28 reported ruptures of 

BGRR fuel during the period 1952-1957. These all occurred with the natural uranium fuel. There was 

one rupture of a uranium oxide (U3O8) sample that was being irradiated for the radioiodine-production 

program. Aside from argon-41, iodine-131 was the most important radionuclide that would contribute 

to a potential dose that was discharged to the atmosphere from the BGRR. Bromine-82 and iodine133 were released in somewhat larger concentrations.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 20 of 106

Over 2414 fuel elements, generated in a 12-year period, were shipped from the canal. The 

contaminated water, filter media, and back flushing from the ion exchange columns were pumped to 

the storage tanks at the Waste Concentration Facility, Building 811.

By 1972 the canal was finally drained, cleaned with detergents and water, and covered with concrete 

slabs for shielding.

Building 704 – 

The BGRR was shut down in 1968 and partially decommissioned in 1972. The exhaust ducts from 

the reactor and to the stack have been sealed from the fans. The fans remain in their cells and are 

contaminated. The intake duct, exhaust duct, and fans are grossly contaminated from fuel failures 

that occurred with the natural uranium fuel. 

Building 708 – 

Radioactive hot air was filtered and discharged from the reactor through underground concrete ducts 

located between the reactor building 701, the instrument house building 708, and the fan house, 

building 704. The north and south exit air ducts were contaminated as a result of fuel ruptures with 

the original natural uranium fuel. Due to decay over some 30 years since use of uranium metal fuel, 

the major remaining fission products should be cesium-137 and strontium-90. On August 18, 1988, a 

radiological survey was conducted on each exhaust duct upstream of the filters. Results indicated 

20,000 to 30,000 dpm in the ducting. This indicates 40 to 70 times higher than the unrestricted limit of 

500 dpm. Filter elements and exhaust coolers are still in place.

2.4.2

The BGRR capacity was replaced and surpassed in 1965 by the High Flux Beam Reactor (HFBR). 

The HFBR is a 30 MW thermal heavy water moderated nuclear research reactor. The reactor is 

designed to provide intense beams of neutron radiation to be used in the study of a wide variety of 

neutron scattering research projects and to allow for the radiating of materials close to the reactor 

core by insertion through special tubes positioned near the reactor core. The HFBR is fueled with 

fully enriched uranium and aluminum alloy, moderated and cooled by heavy water. Initial criticality 

was achieved on October 31, 1965. The HFBR was originally designed to operate at 40 MW. In 1979 

a newer fuel design was employed which provided increased uranium loading in anticipation of an 

overall facility upgrade. In 1982, several major modifications were completed that allowed operation 

at powers up to 60 MW. 

High Flux Beam Reactor (HFBR), Building 750

In 1990, as a result of questions raised about the ability of natural circulation to cool the reactor during 

certain emergencies, reactor operations were limited to 30 MW. Heavy water flowing in the core is 

exposed to a dense neutron field which activates the deuterium atoms in the water to produce tritium. 

The primary mechanism by which tritium is transferred from the interior coolant system to the 

atmosphere is depressurization of the reactor vessel and evaporative losses during maintenance and 

refueling operations. During a scheduled maintenance shutdown in 1997, a leak in HFBR’s spent fuel 

storage pool was discovered. In November 1999, the Secretary of Energy made a decision to 

permanently close the HFBR. 

The HFBR Complex includes the following buildings:

Building 750 – 

Building 750 is a domed cylindrical structure which contains the reactor and almost all associated 

process systems. The reactor, its beam lines and laboratories are located on the second main level 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 21 of 106

known as the Experimental Level. The top most level, the Operations Level, contains the Control 

Room and the Instrument Maintenance Shops.

Building 751 – 

In 1980 the Cold Neutron Facility (CNF) was installed at beam line H9. This facility provided 

extremely low energy neutrons by moderating thermal neutrons with liquid hydrogen. Cooling of the 

liquid hydrogen was provided by liquid helium cooled by a large compressor installed in building 751. 

Building 705 – 

Building 705 is a 100 meter stack that provides the path for ventilation exhaust from the HFBR. A 30 

inch underground duct exhausts air from Building 750 ventilation blowers to particulate and iodine 

filters and then up through the stack. Stack effluent, downstream of the filters is monitored from the 

Stack Monitoring Facility in Building 715. 

2.4.3

The Medical Department (Building 490) was opened in 1956 and currently conducts research in body 

composition and imaging, and Boron Neutron Capture Therapy (BNCT) in the building. Active clinical 

research involving human volunteers is conducted in the Clinical Research Center and its satellite 

facilities. Animal research is conducted in the Brookhaven Lab Animal Facility (BLAF). Bench top 

laboratory research is conducted in support of the various programs. Studies involving dispersible 

radioisotopes (used primarily as tracers) are carried out in the building.

Medical Research Center (MRC), Building 490 

There are several other facilities housed within the building, among them the Whole Body Neutron 

Irradiation Facility, the Prompt Gamma Neutron Facility, the Partial Body Neutron Activation Facility, 

the Inelastic Neutron Scattering Facility, the Whole Body Counter, and two SPECT cameras. Some of 

these facilities contain radiation sources, e.g., PuBe sources for the Whole Body Neutron Irradiation 

Facility (WBNIF) are stored in a vault below the facility and raised into the facility when in use. In 

addition, there are many fume hoods and other apparatus typically found in research labs. 

2.4.4

The Medical Research Reactor (BMRR), Building 491, is the first nuclear reactor built exclusively for 

medical and biological research and came on-line on March 15, 1959 and operated until October 

2000. It is an integral part of the Medical Research Center and is housed in a 60 foot diameter steel 

and concrete structure adjacent to the Medical Research Center (Building 490). It is fueled with 

enriched uranium, moderated and cooled by light water, and is operated intermittently at power levels 

up to 3 MW (thermal). The reactor is capable of operating for short periods of time at powers up to 5 

MW. Air from the interior of the containment building is used to cool the neutron reflector surrounding 

the core of the BMRR reactor vessel. When air is drawn through the reflector, it is exposed to a 

neutron field which causes the natural argon gas in the air to become radioactive (as argon-41). Due 

to a reduction of research funding, the BMRR conducted its last run on December 2000, and transition 

and stabilization activities began in 2001. 

Brookhaven Medical Research Reactor, Building 491

The research function of the reactor was to provide radiation beams of known, controllable character 

and strength at the designed sample and treatment locations. In 1997, its primary use was to provide

neutron beams for Boron Neutron Capture Therapy (BNCT) research and patient treatment. Testing 

of neutron capture compounds and techniques was ongoing. 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 22 of 106

2.4.5

Building 906 was constructed in 1981 and houses two state-of-the-art Positron Emission Tomographic 

Scanners for use in basic research in the fields of neurology, nuclear medicine and neurosciences.

Positron Emission Tomography, Building 906

Radiopharmaceuticals labeled with short-lived positron emitters (carbon-II, fluorine-18, nitrogen-13 

and oxygen-15) are used for physiological and biochemical research.

In 1987 the research group acquired a new PET scanner and another in 1997. The name-changes 

for Building 906 have been evolutionary in that today it is called the Pet Imaging Laboratory.

2.4.6

The 902 Complex was originally built for the Cosmotron proton accelerator and experimental facilities. 

Some of the areas date back to Camp Upton. The Cosmotron was a very low intensity accelerator.

Cosmotron, Building 902

The balance of the high bay area of 902 was used for research and development of superconducting 

magnets. Eventually the Medical Department moved their equipment out and the high bay was 

devoted entirely to the development and fabrication of superconducting magnet.

The operational Chronology of the Building 902 High Bay is:

1949–1968 Cosmotron

1968–1980 Radiation effects facility in the south portion of the high bay

1968–present Superconducting magnet development and fabrication and other equipment 

assembly

2.4.7

The Alternating Gradient Synchrotron (AGS) achieved full energy in 1960 and was capable of 

accelerating protons to energies up to 30 GeV and heavy ion beams to 15 GeV/amu. When such 

accelerated particles collide and interact with nuclei, part of their energy is transformed into new 

particles, which fly off from the target nuclei. These particles are then detected or their paths are 

made visible by devices such as scintillation counters, Cerenkov counters, bubble chambers, spark 

chambers, and photographs. The AGS has been continuously upgraded and it now provides the 

highest intensity of protons per pulse and the highest intensity of polarized protons per pulse than any 

other synchrotron in the world. The AGS also accelerates heavy ion beams for experimental use as 

well as for the Relativistic Heavy Ion Collider (RHIC). 

Alternating Gradient Synchrotron, 900 Series Buildings 

The AGS consists of three different accelerating systems: the 200 MeV LINAC which is the source of 

the protons housed in Building 930, the 200 meter circumference rapid cycling Booster synchrotron 

located in Buildings 914 and 942, and the 800 meter AGS main ring designated as Building 913. The 

Tandem Van de Graaff accelerator, which is a separate facility, provides heavy ions for the Booster 

ring via a beam transfer line. The particles which are accelerated by the AGS facility are extracted 

into and stopped in various experimental areas located around the ring, or sent via a transfer line to 

the RHIC. Building 912 was the original experimental target area. The asphalt pad on the inside of 

the AGS ring adjacent to Building 912 was an experimental area in the 1960's with two secondary 

beam lines. Since that time activated components have been stored and repaired in the area. Figure 

2-3 shows the buildings associated with the AGS Complex. 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 23 of 106

Figure 2-3. The AGS Complex. 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 24 of 106

The AGS Linear Accelerator (LINAC), Building 930 and Associated Facilities:

The LINAC was designed and built in the late 1960's as part of a major upgrade to the AGS facility. 

The Linear accelerator tunnel is 140 m long and consists of nine large accelerators. From the end of 

the LINAC a 17 meter long spur tunnel injects the beam into the Booster accelerator. Another highenergy beam transport (HEBT) line extends 95 meters to connect with the AGS. Another spur tunnel 

provides beam to the Brookhaven Linac Isotopes Producer (BLIP) facility, the Chemistry Linac 

Isotopes Facility (CLIF), the Radiation Effects Facility, and the Neutral Beam Test Facility, and a 

beam stop. Only the BLIP facility is presently used and it has recently undergone a significant 

upgrade to increase its intensity. The Linac consistently produces a beam of 25 to 35 mA with a 0.5 

ms pulse length or 1.0 × 1014 protons/pulse. Pulses are repeated at the rate of 7.5 per second, but at 

most only 1 in 7 pulses are transported to the Booster with the remainder transported to the BLIP 

facility. 

Relevant buildings directly attached with building 930 are:

Building 930 MER - provides water-cooling services for the Linac complex and contains activated 

water. Associated with it is a large water- cooling tower and a compressor area. 

Building 930 Annex - was added in the 1970's and contains shops and storage areas. 930A was 

added on in the late 1980's as a service building for the Booster project. It contains power supplies 

and computer control equipment for the Booster.

The AGS Accelerator Rings, Buildings 913 and 914 and Associated Facilities:

The AGS tunnel, Building 913 has a circumference of 800 meters and is buried under a 7 meter thick 

berm of earth and concrete. There are 240 large alternating gradient dipole electromagnets, over one 

hundred supporting focus and beam tuning magnets as well as RF accelerating components, vacuum 

systems, water cooling systems, etc., located in the tunnel. Supporting the tunnel operation are the 

ventilation system fan houses and multiple power supply houses. Some of these houses are 

designated 913A through R, while others have a separate plant engineering building number. The 

AGS tunnel (building 913) is interrupted at the Experimental Hall, Building 912, for roughly 65 meters. 

Here the AGS is under heavy concrete shielding blocks. 

The Booster tunnel (Building 942) is directly adjacent to the AGS tunnel and the LINAC. It is 

connected to the LINAC and the Tandem Van de Graaff via the HEBT transfer tunnel. The original 

connecting tunnel is now used to transfer beam from the Booster to the AGS. The Booster tunnel is 

200 meters in circumference and is buried under a 5 meter thick berm. 

The new section of the Booster tunnel was built in 1988. Some radioactive gases were created by 

activation of the ventilating air within the tunnel and experimental areas of the AGS. But, since they 

are short-lived (carbon-11, half-life 20.5 min; nitrogen-13, half-life 10.0 min; oxygen-15, 2.1 min), they 

did not create a measurable radiation level beyond the immediately adjacent on-site area. All of the 

cooling water systems in the AGS and Booster tunnels are activated. This includes systems which 

use the AGS ring as a utility tunnel for piping which goes to the support buildings. In addition, there 

are activated components in the AGS and Booster rings and activated spare parts are stored in the 

facility.

Other Related Buildings:

Building 911 - is the main office for the AGS Department. It was one of the first structures built for the 

accelerator complex and has direct passageways to both the AGS tunnel (building 913) and the AGS 

experimental area (building 912). Building 911 also contains the main control room, the support 

building for the main magnet cooling system, the support building for the Westinghouse main magnet 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 25 of 106

power supply (now the back-up power supply for the ring magnets), a high bay assembly area with 

magnetic field testing shops, and other light electronics, vacuum, and machine shops. There are 

activated components from the AGS ring which have been stored and repaired in the high bay area 

and shop areas of building 911 in posted radiation work areas. Because it also contains a cooling 

system to cool water from the AGS ring, it does contain activated water. 

Building 912 - Building 912 is the designation of the approximately 5 indoor acres of AGS 

experimental floor. It is not one building, but 5 connected structures constructed over a 30 year 

period. The initial portion of Building 912 was started in 1958 to enclose both a portion of the AGS 

ring and the first experiments. 

Within this large area, a majority of the ~400 AGS experiments have been located. The AGS 

experimental floor also contains a section of the AGS ring. Surrounding the beam lines are concrete 

and steel shielding needed to maintain radiation levels to within the required limits outside the lines. 

Shielding around the proton target stations that produce secondary beams is much thicker due to the 

high levels of radiation generated in the targets. This shielding can have significant activation 

contained within the body of the material. Cooling water at these target stations has higher levels of 

activation than the magnet cooling water and is handled by closed cooling systems. Radiation safety 

is provided by security systems that prevent access to these target stations and primary beam lines 

when protons are being extracted to these areas. There are activated components in the beam lines 

including highly activated targets. Steel cutting including activated shielding steel has been performed 

both inside and outside of building 912. Depleted Uranium was used in experimental areas in the 

past, but its use and disposition was tightly controlled. Small radioactive sources (nCi) are used. 

They are inventoried and stored in locked shielding boxes. 

Building 913A-E - AGS ventilation system fan houses were an original part of the AGS ring and 

provided air-conditioning and heating for the tunnel enclosure. Because of their proximity to the 

tunnel they are High Radiation Areas when there is beam in the AGS ring and must be kept secured 

during operations. The cooling water for the air-conditioning system is activated and there is a 

potential for release in the fan house. Air handled in the fan houses is from the AGS tunnel air and 

can also be activated.

Building 914 - Building 914 was the location for the original 50 Mev AGS Linac. This facility contained 

power supplies, water-cooling systems, pumps and compressors, and a motor generator set. During 

the 1970's and 1980's, after the LINAC was decommissioned, this building housed a mechanical 

maintenance group for the AGS division. Activated components were repaired and stored in this 

facility and new components were assembled and cleaned there. 

Building 919 (The g-2/Bubble Chamber and Associated Facilities) - Building 919 was built in 1962 for 

an AGS experimental area and is still in use. The high bay area of 919 housed the 80 inch bubble 

chamber which was in operation until 1978. The bubble chamber had both a cryogenic component 

and a large mechanical component driven by a high-pressure oil hydraulic system. When the bubble 

chamber reached the end of its run, the high bay area and control room were used as shops and 

assembly areas. The cryogenic shop continued to be used to support liquid H2 target in Building 912 

and the other buildings became storage areas. This continued to the late 1980's when the area was 

modified for the g-2 experiment and the water-cooling systems were used to support Booster 

operations. A release of tritium from a gas chronograph detector occurred in 1973 in the technician 

machine shop. The detector held approximately 250 mCi of Tritium. 

Building 920 - Building 920 is the EI0 Power Supply building, and is located above the AGS ring. It 

houses an assortment of power supplies used under different beam configurations, but is presently 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 26 of 106

running only a few supplies. The radiological hazard in Building 920 is radiation from activated 

cooling water, and oil in transformers. 

Buildings 928 (Siemens motor generator house) - Building 928 is directly adjacent to and at one 

corner connected to Building 929. Both buildings share a common basement area. The basement 

area contains support electrical equipment and lubricating equipment for the motor generator set. 

Also in the basement is a mechanical equipment room with water-cooling equipment for power 

supplies in 928 and 929 and for systems in the AGS ring. There were potential radiation hazards from 

activated cooling water.

Building 949 - This new building houses the proton target that produces secondaries for the g-2 ring. 

The water- cooled target is highly activated (>50 rem at contact immediately after the beam has been 

turned off). Building 949 is built into the side of the V-line and has extensive shielding. The beam 

line, which runs between the target building and 919 contains water-cooled magnets and vacuum 

pumps. The cooling water is shared with the target so it is activated. 

Building 960 – The building was the site of a large high-energy physics experimental facility, the 

Seven Foot Bubble Chamber (1972-1979). After the Bubble Chamber was removed from 960, the 

facility was used for storage and to test superconducting materials, in conjunction with a small 

cryogenic helium refrigerator (1979-1993). The testing of superconducting materials moved to 

Building 902 in 1993. From 1993 to 1997, the building was used for storage and then abandonment. 

From a radiological standpoint, the particle beam delivered to the Bubble Chamber was very low 

intensity with respect to causing induced activity in the machine components.

Buildings 912, 919, 975 - The AGS Department has used warehouse space throughout the facility and 

the Laboratory. The AGS also stored materials outside, in the inner mongolia area of the AGS, the 

southwest area, and in the ISABELLE experimental buildings. One of the major issues confronting 

AGS storage is the storage of activated materials. When a part fails in service in the accelerator 

tunnels or the experimental lines it is often replaced with a spare. After a radiation cool down period 

the component is repaired and stored until needed. 

Building 918 (Warehouse) - This building is located adjacent to Building 912. It was built in 1957/58. 

It was a warehouse from the start and always used for that purpose. It was doubled in size in 1962. 

Activated components are stored in it.

Building 196 (Warehouse) - This wood frame building was built in 1942. It was formally abandoned 

and demolished in 1995. The building was used by the AGS department for storage from 1975 until it 

was abandoned in 1991/92. The building housed depleted uranium, lead, activated copper magnet 

coils, vacuum pumps with oil, and other activated components. 

Building 209 (Warehouse) - This warehouse belongs to the S&EP division. The AGS has been 

storing activated components (magnets, jacks/stands, and coils @ <5 mR) in the building since 1995. 

Building 424 (Warehouse) - Was built in 1942 for Camp Upton as its theater. The AGS began storing 

activated components and power supplies with PCB capacitors there from 1973 until it collapsed in 

1996. After the building collapsed it was totally demolished. 

Building 178 (Warehouses) - This building was built in 1922 as a theater/lecture hall. The AGS began 

storing activated components there in the early 1970's. Because the building was wood frame and in 

a "populated" area the levels of activation was low and much of the equipment was new spares. The 

AGS equipment was removed and the building was demolished in 1992.
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2.4.8

In 1972 the Brookhaven LINAC Isotope Producer (BLIP -Bldg. 931B) and the CLIF (931A) were 

constructed on a large earthen mound. The CLIF was operated independently by the Chemistry 

Department until 1977, at which time it was incorporated into the BLIP operation. Building 931C was 

constructed in 1996 as part of a substantial facility upgrade.

Brookhaven LINAC Isotope Producer, Building 931

The BLIP was designed for production of radioisotopes via irradiation in the LINAC beam. It consists 

of a main building which houses a "hot cell" over a shaft which descends approximately thirty feet. 

The shaft allows targets containing various substances to be lowered along a track down to an area 

into which a beam from the LINAC is conducted. Irradiated materials are withdrawn, transported to, 

and processed in, the Target Processing Lab (TPL) (Building 801). 

A smaller structure (931A) houses the control panel for BLIP operations and monitoring, as well as a 

single fume hood and a second LINAC beam access system (currently not in use). 

Building 931C acts primarily as a garage to house the fork truck used to deliver irradiated targets to 

the TPL. 

In the BLIP targets are irradiated with 200 MeV protons. In this process high energy secondary 

neutrons are created. These penetrating neutrons are absorbed in the shielding soil placed around 

the main BLIP containment tank. This leads to some activation of the shielding soil.

Building 946 houses the water-cooling equipment for the BLIP target area. This area was upgraded 

to provide better containment and shielding for this activated water system. 

2.4.9

The Tandem Van de Graaff is housed in Building 901. It consists of two Van de Graaff machines, 

arranged so that they can be used independently or in tandem, each capable of accelerating atomic 

particles to energies of 10 MeV. A special device that changes negative ions to positive ions makes it 

possible to achieve a maximum total acceleration of 30 MeV.

Tandem Van de Graaff, Building 901

Building 901 is divided into laboratories, accelerator vaults and office space. The building is the home 

for research that includes medical radionuclide research, accelerator target development, synthetic 

organic chemistry of radiotracers, and analysis of a wide variety of materials. Areas inside the 

building include the cyclotron area, the RFQ-DTL area, laboratory space associated with the 

Department of Applied Science (DAS), and a mechanical shop area which is currently occupied as a 

storage area. The west end of the building contains the Dynamitron, which has been shut down, and 

a robot room dedicated to the analysis of biological samples. 

Area I -Cyclotrons -Two cyclotrons housed in this area are used for the production of 

radiopharmaceuticals. The radiopharmaceuticals are processed in the adjacent "hot" chemistry 

laboratory suite (Rooms 108-112). The cyclotrons are operated remotely from Room 118. These 

machines are used almost exclusively for producing the short-lived positron-emitting isotopes used in 

the PET program. The radioactivity produced by these cyclotrons is remotely transferred, as liquids or 

gases, from the vaults to the adjacent "hot" lab. The "hot" lab contains preparation and purification 

equipment as well as shielded hoods for the processing of the radioisotopes from the cyclotrons into 

radiopharmaceuticals. 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 28 of 106

Area 2 -RFQ-DTL Accelerator - The BNL RFQ-DTL facility consists of the AccSys RFQ accelerator, 

associated electronics, and other support systems. The areas associated with the accelerator include 

the vault (Room 130), where the accelerator is located and the control room (Room 129) where the 

power supplies and control console are located. Since 1996 this facility has occupied space formally 

used for the 3.5 MeV Van de Graaff.

Area 3 - Laboratory Space - This area is laboratory and office space devoted to DAS projects. Most 

of the space is used for data analysis and storage of materials to be analyzed at the NSLS. The 

former machine shop is used for storage of samples and unused electronic equipment associated with 

the 3.5 MeV Van de Graaff.

Area 4 - 901W Dynamitron and Robot Room 

Most of Building 901 was constructed in 1948 and 1949. It was originally constructed to house the 

cyclotron(s) and the 3.5 MeV Van de Graaff. The 3.5 MeV Van de Graaff was installed in the building 

in 1950. 

- The room in the southwestern corner of building 901W 

is dedicated to the operation of a robot used by the BNL PET program to analyze plasma samples 

taken from subjects during the PET study. The rest of the area in 901W is occupied by the 

Dynamitron.

In 1958 building 901 W was constructed to house the 18" cyclotron which, in 1959, was moved from 

the test shack in the rear of the Cosmotron and reassembled in the new building. The 18" cyclotron 

was capable of accelerating protons to energies of about 3 MeV. During the years 1960 to 1963, this 

accelerator was used to study elastic and inelastic scattering of polarized neutrons. The target 

material for the production of the neutrons was a T(p,n)3He reaction on a target of zirconium tritide. 

The accelerator was capable of consistent beam currents of 100 uA and occasionally reached beam 

currents of 200 uA. 

The 60" cyclotron was installed in 1950 and became operational in 1951. The 60" cyclotron was 

capable of producing 10 MeV protons, 20 MeV deuterons, and 40 MeV alpha particles. It was used to 

produce a variety of radioisotopes for use in medicine, biology, chemistry and physics. Upgrading of 

the 60” cyclotron began in 1964 and was completed in 1968. The upgrade project resulted in the 

cyclotron becoming a variable energy machine capable of producing 35 MeV protons, 23 MeV 

deuterons, 57 MeV He-3 and 46 MeV alpha particles. During the late 1950's and early 1960's, a 

number of radioisotopes were in development and/or production at what was known then as LEAF 

(Low Energy Accelerator Facility) which included the cyclotrons, and the Van de Graaff accelerator. 

The isotopes were: Y-90, Ca-47, Cu-67, Ar-38, Mg-28, I-124, I-133, Sc-47, Po-210, Kr-83m, Ga-68, 

Sm-151, Te-132-I-132 generator, and Mo-99-Tc-99m generator. The major purpose of these isotopes 

was medical research. 

In 1965 the Dynamitron was installed in 901W. It was essentially an electron or positron accelerator 

designed for physics research. In the early years, it was used exclusively for the study of solid state 

physics. It was used as a positron accelerator during the late 1980's into the 1990's to study positron 

reactions. During this time it used a large Na-22 source as the generator of positrons. This source 

was removed when the facility was shut down in 1996. At present the facility is not in operation and 

there are no immediate plans for a restart. All activated items have been removed.

The Van de Graaff was installed in building 901 in 1950. There were some problems during the early 

years and a reconstruction project was completed in 1954. In 1960 the first experiments with a He-3 

beam were begun. The first test of the triton beam was carried out in June of 1968. During 1968 and 

for a few years after, Mg-28 was being produced routinely on the Van de Graaff using the 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 29 of 106

26 Mg(t,p)28 Mg nuclear reaction. During the period of 1968 through the late 1970's the 3.5 MeV Van 

de Graaff was being used to accelerate tritium for production of isotopes as well as basic physics 

studies. In 1985, the Van de Graaff was turned over to the DAS. Some experiments with tritium 

continued until the shut down of the Van de Graaff in 1991.

2.4.10

Building 725 is the National Synchrotron Light Source (NSLS). It is one of the most intense sources 

of x-rays, ultraviolet and infra-red radiation in the world. The beams of radiation are created by 

electrons circulating in two storage rings. The x-ray ring energy is 2.5 GeV and the VUV ring is 800 

MeV. The radiation produced by the circulating electrons travel down 85 independent beamlines to 

targets on a wide variety of experiments. There are also wet labs available for the experimenters, 

storage space, offices and a user-machine shop in the building. 

National Synchrotron Light Source, Building 725

The sole occupant of Building 725, which was constructed in 1981, has been NSLS. In 1987 a major 

Phase II expansion of the facility (office space, beamlines, laboratory and set-up space) was added, 

but the original function of the facility remained the same. Operations of the VUV ring commenced in 

1982 and the x-ray ring in 1984.

Building 729 is the National Synchrotron Light Source, Source Development Laboratory. The building 

houses a linear accelerator coupled to a magnetic "wiggler" which produces an intense x-ray beam. 

2.4.11

The RHIC Complex is shown in Figure 2-4. It is a high-energy particle accelerator that was 

commissioned in 1999 and achieved its first successful operation in the summer of 2000. The RHIC 

is designed to achieve much higher reaction energies by colliding two accelerated beams head-on. 

When the ion beam is traveling at top speed in the AGS, it is taken down another beam line called the 

AGS-To-RHIC (ATR) transfer line. At the end of this line, there is a “fork in the road”, where a 

switching magnet sends the ion bunches down one of two beam lines. Bunches are directed either 

left to the clockwise RHIC ring or right to travel counter-clockwise in the second RHIC ring. From here 

on, the counter-rotating beams are accelerated, as in the Booster and AGS, and then circulate in 

RHIC where they will collide into one another at as many as six interaction points. These intersection 

points are described in Figure 2-5.

Relativistic Heavy ion Collider Complex, Building 1000 Series

The chronology of the RHIC buildings:

1981–present Experimental hall, assigned to BRAHMS experiment

Building 1002 

1982–1992 AGS and ADD storage, some activated components

1990–1992 AGS experiment 813

1981–present Utilities, no usage prior to 1992 

Building 1004A (Service Building)

1992–present RF System testing

1981–present Future experimental area

Building 1004 Open Area

1994–present RF cavities testing for Collider

1996 -present Temporary beamstop for Sextant Test completed during January 1997
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 30 of 106

Figure 2-4. The Relativistic Heavy Ion Collider (RHIC) Complex. 

1994–present Power supplies, Cryogenic Valve Boxes, other system electronics

Building 1004B (Support Building)

1981–present Commissioned in January 1997 as part of the Sextant Test 

Building 1005 Beam Tunnel

1981-present Tech Shops on 1st Floor, Administrative Offices 2nd, 3rd, 4th Floors

Building 1005S

1991–present STAR Detector

Building 1006

1985–1986 Machining of low-level radioactive steel for AGS Experiment 787

1981–1990 Other short term uses, temporary storage
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 31 of 106

Figure 2-5. RHIC ring intersection points. 

1981–present Experimental support, tech shop, utilities, and center room occasional storage

Building 1006A (Service Building)

1981–present Beam tunnel to be commissioned in 1999

Building 1007

1981–1995 Part of tunnel used as a calibration lab for Survey Group

1991–present PHENIX Experiment

Building 1008

1992–1995 Injection magnet fabrication

1988–1991 Uranium plate machining for Experiment DO at FNAL 

1981–present Experimental Support Building, Utilities

Building 1008A

1992–1995 Injection Magnet Group Shop/South Room

1990–1992 AGS/Physics Experiment Prep/South Room

1986–1987 Assembly of Wire Chambers for AGS Experiment 755

1985–1986 Cleaning of Depleted Uranium with TCA for CERN Experiment NA-34

1994–present PHOBOS Experimental

Building 1010

1994–present Future Experimental Hall

Building 1012
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 32 of 106

2.4.12

Building 801, the TPL, was renamed the Hot Laboratory in 1993. The Hot Laboratory officially opened 

January 15, 1951. It consisted of a central laboratory, a fan house, a radioactive-liquid waste tank 

farm, and a liquid radioactive-waste concentration plant. The original purpose of the central facility 

was to provide appropriately shielded areas for research with large amounts of radioactive material. 

The building was designed as two separate but connected structures; a "Hot" (the western portion) 

and a "Cold" (the eastern) side. The “hot” area of the Hot Laboratory included five hot cells, three 

chemical-processing hot cells and three high-level hot cells for handling and processing of relatively 

high levels of radioactivity in gaseous, liquid, or solid form. The cells were maintained at negative 

pressure with respect to their surroundings to minimize the possibility of radioactive releases to the 

building. Each has individual exhaust air-filters as well as a backup filter preceding discharge to the 

BGRR stack.

Hot Laboratory, Building 801 

Currently in the Cold area there are a number of chemistry labs that conduct basic research. On the 

Hot side, radioisotope processing and production (related to the BLIP), and radiopharmaceutical 

research is conducted by Medical Department staff.

There are Metallurgy Hot Cells (the Metallurgical Evaluation Lab-MEL) located in the Semi Works 

Area (Area 56) that were constructed in1958. Metallurgical studies on highly radioactive reactor 

components are performed in this facility by Department of Applied Technology (DAT) personnel.

There are "D" & "F" waste systems (storage tanks and piping) housed in the building basement. The 

"D" waste system holds radioactive non-hazardous liquid waste. The "F" waste system holds nonradioactive, non- hazardous liquids. The Hazardous Waste Management Group of S&EP manages 

that facility.

Some additional special facilities in the building are the TPL (a series of six "hot" boxes in a separate, 

rad-monitored area -Room 66), the MEL, and three inactive "Hot Cells". The 300 ft. stack south of the 

building receives the ventilation from hoods on the hot side. The stack is monitored for radioactivity. 

There was a Co-60 irradiation pool in what is now the TPL (Lab 66). The pool, which was constructed 

of concrete, had depths of 4', 10', 14' where large amounts of sealed Co-60 sources were stored 

under water. There was a drain to "D" waste in the deep portion of the pool. The pool was in 

operation from about 1952 until about 1960. The Co-60 pool was moved, water drained and concrete 

decontaminated and backfilled with sand.

From March of 1952 to June of 1960 there was a program at the Hot Laboratory to produce iodine131 by acid-dissolution of irradiated uranium samples. A separate exhaust system, that included 

NaOH scrubbers and backup charcoal filters, was installed in the hot cell and ducted to a stainless 

steel pipe within the BGRR stack.

There are pneumatic tubes in the roadway connecting the BGRR to Building 801. These lines 

allowed quick transfer of short-lived isotopes from the Reactor to Building 801 where they would be 

used. An interviewee reported that at times the rabbits containing the radioactive samples would 

break or get stuck in these tubes. 

Incidents: 

On 5/15/57 a very serious incident occurred involving the reprocessing of uranium for the Volatility 

Project UF6/BrF3 exploded and a few people were injured, one hospitalized. There was a release of 

about 13 lbs of un-irradiated uranium to the local environment. Damaged occurred to nearby trees, 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 33 of 106

automobiles and building equipment. Environmental sampling of the appropriate areas may identify 

residual uranium contamination.

In August of 1960 some highly radioactive waste (reactor fuel elements) were dissolved in aqua regia 

and stored in plastic containers in the storage vault (Room 51). These containers cracked and leaked 

onto a stainless steel tray containing the plastic containers. The solution was sucked up using an 

aspirator connected to the house vacuum system. The house vacuum system became highly 

radioactive. The main vacuum chamber removed all standing liquid in the lines. However, because 

of the high contamination level and long half-life of the activity, the pipes remain residually 

contaminated. The system was moved to a remote location but some of the piping in the floor is still 

contaminated as a result of this incident. The piping is still contained in the concrete floor and is 

labeled as a radioactive area. 

2.4.13

Building 830 is an office and laboratory complex presently occupied by DAT and RHIC personnel. All 

of the RHIC personnel are housed in a modular structure, which is totally office space. The laboratory 

complex has always been used as an extensive experimental area especially with regard to the use of 

radioactive and hazardous chemicals. This facility houses the Gamma Irradiation Facility (GIF), 2 

inactive hot cells, analytical laboratories, electron microscopy, and associated offices. 

High Intensity Radiation Development Laboratory (HIRDL), Building 830

In 1963 Building 830 commenced operations as the HIRDL. At this time it consisted of the high bay 

area (which houses the 2 hot cells), laboratories, and offices. Cell #1 (preparation Cell) was used to 

fabricate high intensity Cobalt-60 sources for food irradiation programs. Some of the sources were 

then moved to Cell #2 (Irradiation Cell) by means of a transfer canal. In the Irradiation Cell, the 

sources were stored in a 21 ft. deep water filled pit within the cell until needed. Items to be irradiated 

were moved in and out of the Irradiation Cell by means of 2 transfer tunnels under the cell. Sources 

were also kept in the 2 bays for storage and irradiation purposes. 

In 1970 the Low Dosimetry Facility (currently known as the Gamma Irradiation Facility or GIF) was 

added to the northeast end of the main building. This facility included the gamma irradiation pool and 

a machine shop. Sources were stored at the bottom of the pool and samples to be irradiated were 

lowered down via air filled tubes. In 1974 the last of the sources was removed from the pit in the 

Irradiation Cell. Since that time the cells have been used for experiments with lower activity level 

materials, and storage of various radioactive materials and contaminated equipment.

Currently, the two hot cells in Building 830 have very little mission essential work. The Laboratory 

spaces are being used for various experiments, and the GIF currently contains approximately 35,000 

curies of cobalt-60. The GIF is used sporadically for various experiments.

2.4.14

Building 555 is the home of BNL's Chemistry Department. It was built between 1963-1966. The only 

major modification was made in 1996 when the underground Center for Radiation Chemistry 

Research was added. The building contains three floors plus a partial basement and loft. There are 

many extra-large laboratories with at least two fume hoods per lab. There are pipe chases between 

rows of labs to deliver all services to the labs. Transparent glass piping in the basement carries 

waste water from the sinks in the labs to the sewer system. Counting rooms constructed of specially 

selected low-activity materials are used for low background counting. There is a large chemical 

stockroom with separate storage rooms for acids, bases, and organic solvents, as well as a room 

devoted to the washing of laboratory glassware.

Chemistry Department, Building 555
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A large variety of research programs have been conducted during the past 31 years. Many different 

chemicals have been used including inorganic acids and bases, inorganic compounds and metals as

well as a plethora of organic and organometallic compounds. A wide variety of specialized equipment 

and instrumentation is in use in different labs; e.g., there are several laser-based systems in 

operation.

In addition, research with radioactive isotopes has been done since the earliest days. Studies include 

chemical kinetics, hot-atom effects, nuclear chemical studies of nuclear decay schemes, nuclear 

reaction cross-sections, solar neutrino studies and radiocarbon dating. Over the years, the mix of 

research projects in the Department has changed considerably, but suffice it to say that virtually every 

chemical element in the Periodic Table has been used at one time or another in the building.

Some radiological problems associated with the building are:

Disposition of sealed Co-60 irradiation sources in labs 171-173. These sources are contained in 

stainless-steel containers that penetrate the lab floors and are in the ground below. See diagram in 

Appendix 5. There is evidence that one source (40-Curie source strength) is leaking small amounts of 

Co-60 into the water that surrounds the source to provide radiological shielding. Some of this 

contaminated water has gotten into the soil surrounding the buried containment vessel. The amount 

of Co-60 in the water and in the soil is small. (The estimated Co-60 activity remaining in this soil is 

less than 35 nCi.) 

Radioactive materials and counting samples are used and/or stored in many rooms: Labs 225, 227, 

229, 233, 235, 236, 253, 367, 369, 374; and counting rooms 103, 203, 204, 205. Periodically, 

chemical manipulations are done with these materials, in vented hoods that are posted as radioactive 

work areas in accordance with BNL procedures. Occurrence reports from 1990 to present were 

reviewed for building 555. They indicated three separate reportable incidents involving small amounts 

of radioactivity. Two of these, one involving P-32 and the other, U-238, were confined to the 

chemistry labs in which they occurred. The third involved a Kimwipe contaminated with I-124 that was 

inadvertently thrown into the trash, and was subsequently detected in the garbage truck before it left

the BNL site. 

2.4.15

Building 510 was built in 1962 to house the Physics Department personnel and laboratories. It 

consists of approximately 400 offices and 100 laboratories located in an area of about 200,000 ft2

. 

There is a large machine shop to support the research programs as well as a number of smaller 

machine shops. A high bay area is available for assembling equipment for experiments. The 

basement is used predominantly for storage but houses the building support facilities such as 

heating/air-conditioning units and power distribution systems. The northeast parking lot contains 11 

trailers permanently placed for storage of solid materials. These are not climate controlled and no 

chemical or radioactive materials are stored there.

Physics Department, Building 510

The northeast corner of the building houses in the Health Physics Section (Rm 1-136) where there 

are 35 vaults for storage of radioactive materials. There are 9 Type A vaults that are 12" in diameter 

and penetrate 10' into the ground, 6 Type B vaults which are 8" in diameter and penetrate 10' into the 

ground, and 20 Type C vaults which are 8" in diameter and penetrate 3' into the ground. Vault #17 is 

contaminated because it contained a radium source. The source has been placed in a sealed PVC 

pipe and is now stored in Vault #4.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 35 of 106

2.4.16

2.4.16.1 LMFR Support, Building 820

Other Miscellaneous Radiological Research Facilities

Research and Development on Liquid Metal Fuel Reactor (LMFR) technology took place in Building 

820 from 1957 to 1975. A simulated reactor with a core of uranium dissolved in bismuth was studied. 

2.4.16.2 Hot Machine Shop, Building 530

Building 530 functioned as a Hot Machine Shop for the Central Shops Division. In 1975 Central 

Shops evacuated the building and its equipment and function relocated to building 462. The building 

was then utilized as a storage area by Plant Engineering until 1988 at which time it was demolished. 

Operational Chronology:

1947-1975 - Hot Machine Shop for the Central Shops Division

1975-1988 - Storage Area for Plant Engineering

2.4.16.3 Health Physics & Safety Instrumentation, Building 535

Building 535 was constructed between 1962-1963 and was operational in 1964. The building is a 

multi-tenant, multi-functional facility, consisting of a ground floor and a basement. The 

Instrumentation Division, which is responsible for the building, occupies a major part of the ground 

floor for research and development of x-ray and neutron detectors and corresponding electronics. 

These detectors and associated electronics are used at the NSLS, the High Flux Beam Reactor 

(HFBR), the Alternating Gradient Synchrotron (AGS), and other worldwide scientific research facilities. 

A state-of-the-art electronic printed circuit design and fabrication facility and a microelectronics clean 

room facility are unique to the Laboratory. 

In 1978 the NSLS Department moved into the rear section of the basement, occupying that space to 

the present. An underground walkway (tunnel) was constructed in 1986 between the basement of 

Building 535 and the main floor of Building 725 (NSLS Department).

In 1985, Instrumentation Division established the Optical Metrology Laboratory in the southeast 

section of the basement, to conduct research for synchrotron radiation experiments. In 1987, an 

office module was added to the southwest corner of the building.

2.4.17

The Waste Management Facility (WMF) was opened in December 1997 for managing the wastes 

generated from BNL’s research and operation activities. The new WMF replaces the original 

Hazardous Waste Management Facility in its entirety and its design ensures that all storage and 

transfer activities are accomplished inside buildings and on paved and curbed areas A security fence 

around the cleared area and a berm 8 feet high around the perimeter of the radwaste portion of the 

facility ensures that incidental exposures meet the BNL design criteria of 25 mrem/year to nonradiation workers.

Waste Management Facility (WMF), Radwaste Reclamation Building, Building 865

The WMF includes an Operations Building (Bldg. 860), a RCRA Waste Building (Bldg. 855), a 

Radwaste Reclamation Building (Bldg. 865), and a Mixed Waste Building (Bldg. 870). 
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Radwaste Reclamation Building, Building 865 - The Reclamation Building is the primary facility for 

radioactive waste handling, size reduction, and repackaging for subsequent off-site disposal for BNL. 

The goal is to size reduce the radwaste stream generated at BNL. This building receives bulk 

radioactive waste of various sizes and configurations to be disassembled, decontaminated, size 

reduced and packaged for temporary storage prior to shipment off-site. To achieve the goal of waste 

reduction, Bldg. 865 will house a 500 pound per day lead smelter, a trash compactor, and a shredder.

Additional radwaste treatment facilities include: 

Wastewater processing began in 1995. The Tritium Evaporator Facility (BLDG. 802B) was 

constructed to reduce the total amount of tritiated water released to the Peconic River. The other 

room contains fans that exhaust Building 801 out the stack. 

Bldg. 811 was designed in the late 1940's and constructed in 1950 to support the radiological waste 

stream from the Nuclear Engineering Department and Medical Department in the 801 and 701 

complexes. Liquid waste, which contains residual radioactive material and generated on-site, is 

processed at the Waste Concentration Facility (WCF), Building 811. At the WCF, suspended solids 

are removed from the liquid along with a high percentage of radionuclides using a reverse osmosis 

process. The only radionuclide which is not removed during this process is tritium. The tritiated water 

which remains following the waste concentration process is delivered to the Evaporator where it is 

converted to steam and released as an airborne effluent.

The Waste Management Incinerator. The Laboratory incinerates certain low-level radioactive wastes 

at the HWMF incinerator, Building 444.

2.4.18

The Waste Decontamination Facility was constructed in 1957 and occupies the western half of 

Building 650. It replaced the yard decontamination operations previously done at Building 446. 

Decontamination of large pieces continues until the 1980s. The process slowed and is now idle.

Radiological Waste Decontamination Facility, Building 650

The facility consists of an office, showers, decontamination showers, airlock, and high bay. The 

decontamination showers were once used for decontamination of equipment. They drain to the 

basement then to the "D" waste tanks (removed). The high bay houses a lead melter (contaminated 

vent pipe; out-of-service), two vapor blaster talc/water abrasive blasters (internally contaminated; outof-service), one walk in shot blaster (internally contaminated; out-of-service), sample hood, and 

limited storage (one B-25 container holding film badge records at the time of the walk through). 

Already removed were two wet decontamination tanks (a 20' × 4' × 4' top tank used for 

acid/base/detergent wash).

Associated with the Waste Decon Facility is an outside decontamination pad. From 1957 to 1981 the 

20' × 20' pad was used for washing off surface contamination of large pieces of equipment. Drainage 

from the pad used to go to the underground storage tanks (now removed). The surface of the pad 

has low level radiological contamination. About 1994 an asphalt cap has been laid by OER to contain 

the radiological materials. Soil in the vicinity of the pad has been characterized by OER and is in their 

remedial plan (1993 Survey, Cs-137, Eu-152, Eu-154, depth of 0.5 feet, exceeding OER clean up 

goals). The low level contaminants are fixed in the soil and are unlikely to migrate.

From 1959 to 1996, the Plant Engineering portion of the building was used as a laundry facility for 

both contaminated and clean laundry. In 1996, laundry services were subcontracted and the facility is 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 37 of 106

now being used for storage of custodial supplies. 650T has always been used as offices since they 

were first acquired in 1994.

2.5 ISOTOPES OF CONCERN 

Table 2-2 identifies the potential radiological contaminants associated with the specific nuclear 

operations at BNL in terms of isotopes of concern. Due to the wide variety of research activities within 

a single multiple-laboratory building and a significant number of programs that handle and/or create 

practically all of the isotopes of the elements in the Periodic Table, the isotopes of concern are 

identified forensically from the airborne radionuclide releases by the facility. This method is 

considered adequate for the purpose of this Section 2 of the BNL site profile, that is, to provide 

background information only. 

In the absence of any measurements or studies, NIOSH guidance requires the use of default solubility 

classes and particle size values from the International Commission on Radiological Protection 

(NIOSH 2002, pp. 15, 16).

Table 2-2. Area information and parameters. 

Section Name Description Period

Building 

No. Radionuclide

2.4.1 Brookhaven Graphite 

Research Reactor 

(BGRR)

Uranium fueled Graphite 

moderated and reflected Reactor

28 MW

1950-1969 701 Ba-140

Ce-144

Cs-137

La-140

Nb-95

Ru-103

Ru-106

Sr-89

Sr-90

Zr-95

U-235

U-238

U-234

Pu-239

2.4.1 BGRR fuel canal BGRR fuel storage 1950-1972 709 Same as 

above

2.4.1 BGRR exhaust fan 

house

Highly contaminated ducting and 

fans

1950-1972 704 Cs-137

Sr-90

2.4.1 BGRR exhaust filter 

house

Contaminated HEPA filter 

elements

1950- 

unknown

708 Cs-137

Sr-90

2.4.2 High Flux Beam 

Reactor (HFBR)

30-60 MW thermal heavy water 

moderated research reactor

1965-1999 750 Be-7

Br-77

Br-82

Cs-137

H-3

I-126

Mn-54

Xe-133

Xe-135

2.4.2 HFBR Stack 100 m stack and houses iodine 

filters

1965-1999 705 I-131

I-132

I-133

I-134

I-135

Table 2-2 (Continued). Area information and parameters. 

Section Name Description Period

Building 

No. Radionuclide

2.4.2 Cold Neutron Facility 

(CNF)

Produce extremely low energy 

neutron by moderate neutron with 

liquid hydrogen

1980-

unknown

751

2.4.3 Medical Research Medical Department 1959-present 490
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 38 of 106

Center (MRC)

2.4.4 Brookhaven Medical 

Research Reactor 

(BMRR)

Enriched U fueled, light water 

moderated and cooled reactor

1959-2000 491 Ar-41

Al-26

As-76

Ba-128

Ba-140

Br-82

Ce-141

Ce-144

Co-60

Fe-59

Hg-203

I-124

I-131

I-133

La-140

Mo-99

Na-24

Sb-122

Sc-46

Se-75

Sr-91

Tc-99m

Ti-44

Xe-133

Xe-135

Zn-65

Zn-69m

2.4.5 PET Imaging 

Laboratory

Positron Emission Tomographic 

Scanners for nuclear medicine 

and neurosciences researches

1977-

present

906 C-11

N-13

O-15

F-18

Ge-68

2.4.6 Cosmotron Proton accelerator 1949-1968 902 C-11

N-13

O-15

2.4.7 Alternating Gradient 

Synchrotron (AGS) 

A large accelerator capable of 

accelerating protons to energies 

up to 30 GeV and heavy ion 

beams to 15 GeV/amu

1960-present 913 C-11

N-13

O-15

Table 2-2 (Continued). Area information and parameters. 

Section Name Description Period

Building 

No. Radionuclide

2.4.7 200 MeV LINAC Pre-accelerator and linear 

accelerator providing proton 

source

1967-present 930 H-3

2.4.7 LINAC cooling support 

service

Provides water-cooling services 

for the LINAC complex

1967-present 930 MER

2.4.7 AGS Booster Tunnel Accumulator booster ring 1960-present 942

2.4.7 AGS Experiment Area Experimental target area for the 

AGS ring and later contains a 

proton beam switchyard system

1960-present 912 Radiation 

sources and 

activated 

targets

2.4.7 AGS ventilation system 

fan houses

Ventilation exhaust fans 1960-present 913A-E

2.4.7 AGS LINAC Contains 50 MeV AGS LINAC 1960s 914 H-3

2.4.7 E10 Power Supply 

Building

Houses an assortment of power 

supply

1971-present 920

2.4.7 Siemens motor 

generator house

Electrical equipment and 

lubricating equipment for the 

motor generator

1960-present 928,929

2.4.7 AGS Department Office, field testing shops, 

machine shops, high bays.

1960-present 911 Radiation 

sources and 

activated 

targets
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 39 of 106

2.4.7 g-2/Bubble Chamber Contains the 80” Bubble 

Chamber

1962-1978 919 H-3

2.4.7 Proton Target Proton target that produces 

secondaries for the g-2 ring

1962-present 949

2.4.7 7’ Bubble Chamber High energy physics 

experimental facility

1972-1979 960 H-3

2.4.7 AGS Warehouses Warehouses 1960-present 975,918,19

6,209,424,

178

2.4.8 Brookhaven LINAC 

Isotope Producer 

(BLIP)

Production of radioisotopes via 

irradiation in the LINAC beam 

(200 Mev Protons)

Carbon-11, Fluorine-18, nitrogen13 and oxygen-18

1961-1962 931B As-72

As-74

Be-7 

Co-57

Co-58

Co-60

Cs-132

Cs-137

H-3

Mn-54

Na-22

O-15

Sc-46

2.4.9 Tandem Van de Graaff Van de Graaff accelerators

60” proton Cyclotron producing 

10-MeV protons, 20-MeV 

deuterons and 40-MeV alpha 

particles for medical radionuclide 

research

1970 -

present

901 Yt-90

Ca-47

Cu-67

Ar-38

Mg-28

I-124

I-133

Sc-47

Po-210

Kr-83m

Ga-68

Sm-151

Te-132

I-132

Na-22

H-3

Mo-99

Tc-99m

2.4.10 National Synchrotron 

Light Source (NSLS)

Produce intense sources of xrays, ultraviolet and infra-red 

radiation

1981-present 725

2.4.10 NSLS Development 

Laboratory

Houses a linear accelerator 1981-present 729

Table 2-2 (Continued). Area information and parameters. 

Section Name Description Period

Building 

No. Radionuclide

2.2.11 Relativistic Heavy Ion 

Collider (RHIC) 

Complex

High-energy particle accelerator 

and collider

BRAHMS experimental hall

RHIC service building

Open area (future experimental 

area)

RHIC support building

Beam tunnel

Offices and tech shops

STAR Detector

Service building

Beam tunnel

1999-present

1981-present

1981-present

1996-present

1994-present

1981-present

1981-present

1985-present

1981-present

1981-present

1988-present

1985-present

1002

1004A

1004

1004B

1005

1005S

1006

1006A

1007

1008

1008A
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 40 of 106

PHENIX experiment

Experimental support building

PHOBOS experiment

Future experimental hall

1994-present

1994-present

1010

1012

2.4.12 Hot Laboratory Target Processing Laboratory 1951-present 801 As-72

As-74

Be-7

Co-57

Co-58

Co-60

Cs-132

Cs-137

H-3

Mn-54

Na-22

Au-199

Bi-213

Br-77

Br-82

Se-75

V-48

I-126

Se-75

Ge-69

2.4.13 High Intensity Radiation 

Development 

Laboratory (HIRDL)

Radioactive sources fabrication, 

irradiation cells and since 1970, 

the Gamma Irradiation Facility 

(GIF) 

1951-present

GIF 1970-

present

830 Co-60

Cs-137

2.4.14 Chemistry Department Studies from chemical kinetics, to 

hot-atom effects, to nuclear 

chemical studies of nuclear 

decay schemes and nuclear 

reaction cross-sections, to solar 

neutrino studies and radiocarbon 

dating.

1966-present 555 H-3

2.4.15 Physics Department laboratories, office space, a 

machine shop and basement 

area

1962-present 510

2.4.16.1 LMFR Support Facility Research and development work 

for the Liquid Metal Fuel Reactor

1957-1975 820

2.4.16.2 Hot Machine Shop Hot Machine Shop 1947-1975 530

2.4.16.3 Instrument Division Research and development of xray and neutron detectors and 

corresponding electronics, which 

are used at the NSLS, the HFBR, 

the AGS, and other worldwide 

scientific research facilities.

1964-present 535

2.4.17 Radwaste Reclamation 

Building

Located within the Waste 

Management facility (WMF) for 

processing of radwaste.

1997-present 865

Table 2-2 (Continued). Area information and parameters. 

Section Name Description Period

Building 

No. Radionuclide

2.4.17 Tritium Evaporator 

Facility

Reduce tritium release by 

evaporation

1995-present 802B I-133

Rb-83

Rb-86

Be-7

Co-56

Co-57

Co-58

Co-60

Cs-137

H-3

Mn-54

Na-22
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 41 of 106

Se-75

Zn-65

2.4.17 Waste Management 

Incinerator

Low-level Waste incinerator 1995-present 444 H-3 

Sr-85

I-125

Co-57

Sc-47

Zn-65

C-14

P-32

S-35

Fe-59

Sn-117m

Sn-113

Cr-51

Be-7

2.4.18 Radiological Waste 

Decontamination 

Facility

Decontamination of radiological 

waste

1959-1996 650

2.6 MAGNITUDE OF SITE ACTIVITY

Table 2-3 intends to provide a perspective of the magnitude of the nuclear operations at BNL. The 

Site Profile preparation guide (ORAU 2005) requires the magnitude of the operations be expressed in 

radioactivity (Ci) of the isotope of concerns. The same information is also required in Section 5, 

Internal Dosimetry. Hence, radioactivity values in Section 5 should be used for dose reconstruction. 

In cases where available information does not allow the estimation of the radioactivity at the time of 

the operations, the activity fraction of each isotope is listed. Activity fractions listed in the table are the 

ratios of the activity of the individual isotope released and the total activity released of all the isotopes. 

In cases where numeric data is not available, the table provides the descriptive text that gives a 

perspective on the magnitude of the radiological impact on personnel. 

Table 2-3. Magnitude of Site Activity. 

Section Name Description Period

Building 

No.

Radioactivity

Radionuclide

Activity 

Fraction

2.4.1 Brookhaven 

Graphite Research 

Reactor (BGRR)

Uranium fueled Graphite 

moderated and reflected 

Reactor

28 MW

1950-1969 701 Ba-140 1.70E-01

Ce-144 1.08E-01

Cs-137 7.19E-03

Table 2-3 (Continued). Magnitude Site Activity. 

Section Name Description Period

Building 

No.

Radioactivity

Radionuclide

Activity 

Fraction

La-140 1.95E-01

Nb-95 1.62E-01

Ru-103 7.28E-02

Ru-106 6.09E-03

Sr-89 1.17E-01

Sr-90 7.78E-03

Zr-95 1.53E-01

U-235 8.35E-08

U-238 1.81E-06

U-234 1.81E-06

Pu-239 2.86E-04

2.4.1 BGRR fuel canal BGRR fuel storage 1950-1972 709 Same as 

above

Same as 

above

2.4.1 BGRR exhaust fan 

house

Highly contaminated 

ducting and fans

1950-1972 704 Cs-137

Sr-90
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 42 of 106

2.4.1 BGRR exhaust filter 

house

Contaminated HEPA filter 

elements

1950-

unknown

708 Same as 

above

2.4.2 High Flux Beam 

Reactor (HFBR)

30-60 MW thermal heavy 

water moderated research 

reactor

1965-1999 750 Be-7 3.41E-08

Br-77 3.14E-05

Br-82 7.31E-06

Cs-137 2.5E-08

H-3 1.00E+00

I-126 4.79E-09

Mn-54 1.46E-09

Xe-133 4.06E-07

Xe-135 4.77E-06

2.4.2 HFBR Stack 100 m stack and houses 

iodine filters

1965-1999 705 I-131

I-132

I-133

I-134

I-135

2.4.2 Cold Neutron 

Facility (CNF)

Produce extremely low 

energy neutron by 

moderate neutron with 

liquid hydrogen

1980-

unknown

751

2.4.3 Medical Research 

Center (MRC)

Medical Department 1959-present 490

2.4.4 Brookhaven Medical 

Research Reactor 

(BMRR)

Enriched U fueled, light 

water moderated and 

cooled reactor

1959-2000 491 Ar-41 1.00E+00

Al-26 5.45E-12

As-76 2.12E-07

Ba-128 8.38E-08

Ba-140 7.03E-08

Br-82 3.88E-06

Ce-141 8.20E-11

Ce-144 6.40E-10

Co-60 1.19E-09

Fe-59 1.73E-08

Hg-203 2.66E-08

I-124 8.51E-09

I-131 1.50E-08

Table 2-3 (Continued). Magnitude of Site Activity. 

Section Name Description Period

Building 

No.

Radioactivity

Radionuclide

Activity 

Fraction

I-133 1.64E-07

La-140 3.72E-07

Mo-99 6.94E-11

Na-24 1.04E-07

Sb-122 2.17E-10

Sc-46 9.68E-09

Se-75 9.19E-11

Sr-91 1.50E-07

Tc-99m 2.64E-08

Ti-44 5.81E-08

Xe-133 4.59E-08

Xe-135 4.82E-07

Zn-65 8.78E-09

Zn-69m 6.49E-10

2.4.5 PET Imaging 

Laboratory

Positron Emission 

Tomographic Scanners for 

nuclear medicine and 

neurosciences researches

1977- 

present

906 C-11

N-13

F-18

O-15

Ge-68

2.4.6 Cosmotron Proton accelerator 1949-1968 902 C-11

N-13

O-15
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 43 of 106

2.4.7 Alternating Gradient 

Synchrotron (AGS) 

A large accelerator 

capable of accelerating 

protons to energies up to 

30 GeV and heavy ion 

beams to 15 GeV/amu

1960-present 913 C-11

N-13

O-15

Table 2-3 (Continued). Magnitude of Site Activity. 

Section Name Description Period

Building 

No.

Radioactivity

Radionuclide

Activity 

Fraction

2.4.7 200 MeV LINAC Pre-accelerator and linear 

accelerator providing 

proton source

930

2.4.7 LINAC cooling 

support service

Provides water-cooling 

services for the LINAC 

complex

930 MER

2.4.7 AGS Booster Tunnel Accumulator booster ring 942 C-11

N-13

O-15

2.4.7 AGS Experiment 

Area

Experimental target area 

for the AGS ring and later 

contains a proton beam 

switchyard system

1960-present 912

2.4.7 AGS ventilation 

system fan houses

Ventilation exhaust fans 1960-present 913A-E

2.4.7 AGS LINAC Contains 50 MeV AGS 

LINAC 

1960s 914 H-3

2.4.7 E10 Power Supply 

Building

Houses an assortment of 

power supply

1971-present 920

2.4.7 Siemens motor 

generator house

Electrical equipment and 

lubricating equipment for 

the motor generator

1960-present 928,929

2.4.7 AGS Department Office, field testing shops, 

machine shops, high bays.

1960-present 911

2.4.7 g-2/Bubble

Chamber

Contains the 80” Bubble 

Chamber

1962-1978 919 H-3

2.4.7 Proton Target Proton target that 

produces secondaries for 

the g-2 ring

1962-present 949

2.4.7 7’ Bubble Chamber High energy physics 

experimental facility

1972-1979 960

2.4.7 AGS Warehouses Warehouses 1960-present 975,918,

196,209,

424,178

2.4.8 Brookhaven LINAC 

Isotope Producer 

(BLIP)

Production of 

radioisotopes via 

irradiation in the LINAC 

beam (200 Mev Protons)

Carbon-11, Fluorine-18, 

nitrogen-13 and oxygen18

1961-1962 931B As-72 1.33E-07

As-74 7.24E-10

Be-7 1.60E-06

Co-57 3.56E-08

Co-58 1.44E-07

Co-60 1.40E-08

Cs-132 1.43E-09

Cs-137 2.09E-09

H-3 1.16E-04

Mn-54 1.66E-07

Na-22 6.54E-09

O-15 1.00E+00

Sc-46 2.88E-09

Co-56 1.86E-07

Ga-68 1.38E-08

Ge-69 2.01E-09

I-126 4.17E-08

Xe-127 4.17E-08

Zn-65 5.49E-08

2.4.9 Tandem Van de 

Graaff 

Van de Graaff 

accelerators

60” proton Cyclotron 

producing 10-MeV 

protons, 20-MeV 

1970 -

present

901 Yt-90

Ca-47

Cu-67

Ar-38

Mg-28
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 44 of 106

deuterons and 40-MeV 

alpha particles for medical 

radionuclide research

I-124

I-133

Sc-47

Po-210

Kr-83m

Ga-68

Sm-151

Te-132

I-132

Na-22

H-3

Mo-99

Te-99m

2.4.10 National 

Synchrotron Light 

Source (NSLS)

Produce intense sources 

of x-rays, ultraviolet and 

infra-red radiation

1981-present 725

Table 2-3 (Continued). Magnitude of Site Activity

Section Name Description Period

Building 

No.

Radioactivity

Radionuclide

Activity 

Fraction

2.4.10 NSLS Development 

Laboratory

Houses a linear 

accelerator

1981-present 729

2.4.11 Relativistic Heavy 

Ion Collider (RHIC) 

Complex

High-energy particle 

accelerator and collider

BRAHMS experimental 

hall

RHIC service building

Open area (future 

experimental area)

RHIC support building

Beam tunnel

Offices and tech shops

STAR Detector

Service building

Beam tunnel

PHENIX experiment

Experimental support 

building

PHOBOS experiment

Future experimental hall

1999-present

1981-present

1981-present

1996-present

1994-present

1981-present

1981-present

1985-present

1981-present

1981-present

1988-present

1985-present

1994-present

1994-present

1002

1004A

1004

1004B

1005

1005S

1006

1006A

1007

1008

1008A

1010

1012

2.4.12 Hot Laboratory Target Processing 

Laboratory

1951-present 801 As-72 2.29E-04

As-74 9.8E-04

Be-7 5.23E-04

Co-57 2.11E-04

Co-58 1.79E-03

Co-60 2.39E-04

Cs-132 9.73E-05

Cs-137 2.06E-04

H-3 8.74E-05

Mn-54 5.80E-05

Na-22 5.33E-06

Au-199 4.79E-03

Bi-213 2.14E-02

Br-77 8.97E-01

Br-82 1.54E-02

Se-75 4.94E-02

V-48 6.72E-04

I-126 2.84E-03

Se-75 6.76E-10

Ge-69 4.56E-03

2.4.13 High Intensity 

Radiation 

Development 

Laboratory (HIRDL)

Radioactive sources 

fabrication, irradiation cells 

and since 1970, the 

Gamma Irradiation Facility 

(GIF) 

1951-present

GIF 1970-

present

830 Co-60
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 45 of 106

Cs-137

2.4.14 Chemistry 

Department

Studies from chemical 

kinetics, to hot-atom 

effects, to nuclear 

chemical studies of 

nuclear decay schemes 

and nuclear reaction 

cross-sections, to solar 

neutrino studies and 

radiocarbon dating.

1966-present 555

2.4.15 Physics Department laboratories, office space, 

a machine shop and 

basement area

1962-present 510

2.4.16.1 LMFR Support 

Facility

Research and 

development work for the 

Liquid Metal Fuel Reactor

1957-1975 820

2.4.16.2 Hot Machine Shop Hot Machine Shop 1947-1975 530

Table 2-3 (Continued). Magnitude of Site Activity. 

Section Name Description Period

Building 

No.

Radioactivity

Radionuclide

Activity 

Fraction

2.4.16.3 Instrument Division Research and 

development of x-ray and 

neutron detectors and 

corresponding electronics, 

which are used at the 

NSLS, the HFBR, the 

AGS, and other worldwide 

scientific research 

facilities.

1964-present 535

2.4.17 Radwaste 

Reclamation 

Building

Located within the Waste 

Management facility 

(WMF) for processing of 

radwaste

1997-present 865

2.4.17 Tritium Evaporator 

Facility

Reduce tritium release by 

evaporation

1995-present 802B I-133 1.48E-05

Rb-83 1.08E-05

Rb-86 4.43E-05

Be-7 2.40E-04

Co-56 1.51E-06

Co-57 1.52E-05

Co-58 9.13E-06

Co-60 1.21E-05

Cs-137 2.68E-05

H-3 9.99E-01

Mn-54 9.62E-06

Na-22 6.8E-07

Se-75 1.07E-06

Zn-65 2.24E-04

2.4.17 Waste Management 

Incinerator

Low-level Waste 

incinerator

1995-present 444 H-3 2.66E-02

Sr-85 4.83E-05

I-125 3.14E-02

Co-57 4.83E-06

Sc-47 1.21E-05

Zn-65 1.69E-02

C-14 7.00E-01

P-32 1.93E-02

S-35 1.06E-01

Fe-59 2.41E-03

Sn-117m 2.41E-03

Sn-113 6.35E-03

Cr-51 2.66E-02

Be-7 6.16E-02

2.4.18 Radiological Waste 

Decontamination 

Facility

Decontamination of 

radiological waste

1959-1996 650
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 46 of 106

2.7 MAJOR INCIDENTS

Table 2-4 describes major site incidents that might have significant potential for internal or external 

exposure to personnel.

Table 2-4. Major incidents. 

Facility/building Year Description

BGRR, 

Building 701

1952-57 During the early years of operation of the BGRR, fuel failures occurred which 

resulted in radioactive materials being released to the air stream that cooled 

the reactor. There were 28 reported ruptures of BGRR fuel during the period 

1952-1957. These all occurred with the natural uranium. There was one 

rupture of a uranium oxide (U3O8) sample that was being irradiated for the 

radioiodine-production program. Aside from argon-41, iodine-131 was the 

most important radionuclide that would contribute to a potential dose that 

was discharged to the atmosphere from the BGRR. Bromine-82 and iodine133 were released in some what larger concentrations.

Hot Laboratory,

Building 801 

5/15/1957 There was a very serious incident involving the Volatility Project where 

uranium reprocessing was the objective. There was an explosion of 

UF6/BrF3 and a few people were injured, one hospitalized. There was a 

release of about 13 lbs of un-irradiated uranium to the local environment. 

Apparently, the UF6 combines with moisture readily to form an oxide. The 

BrF3 is very corrosive and damaged nearby trees and automobiles and 

equipment in the Building. 

8/1960 There was some highly radioactive waste reactor fuel elements dissolved in 

aqua regia stored in plastic containers in the storage vault (Room 51). 

These containers cracked and leaked onto a stainless steel tray containing 

the plastic containers. It was decided to suck up the solution with an 

aspirator connected to the house vacuum system. The house vacuum 

system became very highly radioactive. The main vacuum chamber 

removed all standing liquid in the lines. However, because of the high level 

and long half-life of the activity, the pipes remain residually contaminated. 

Thereafter the system was moved to a remote location but some of the 

piping in the floor is still contaminated as a result of this incident. The piping 

is still contained in the concrete floor and is labeled as a radioactive area. 

HFBR,

Building 750

1997 During a scheduled maintenance shutdown in 1997, a leak in HFBR’s spent 

fuel storage pool was discovered. In November 1999, the Secretary of 

Energy made a decision to permanently close the HFBR.

2.8 RADIOLOGICAL ACCESS CONTROLS

Radiological access controls are exercised locally at the site where activities involving radioactive 

materials are performed. Radiological Control Areas are established for each specific laboratory, 

facility or building in which nuclear operations were conducted at the BNL site. Entry to those areas 

was controlled (i.e., hand-and-shoe counters, dosimeters, and training requirements). 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 47 of 106

REFERENCES

Deputy Director, Reactor, Safety and Security, Review of Potential Environmental Release Points, 

Draft, Directors Office Building 701/BGRR, Brookhaven National Laboratory, 1997.

Reactor Division Manager, Review of Potential Environmental Release Points, Department/Division -

Reactor Building/Facility -750/HFBR, 707A/Pumphouse, 707B/Water Treatment House & 

Cooling Towers, Brookhaven National Laboratory, 1997.

Medical Dept. Chairman, Review of Potential Environmental Release Points, Revision 1, Medical 

Department/490, Brookhaven National Laboratory, 1997. 

Reactor Division Manager, Review of Potential Environmental Release Points, Reactor Division, 

Building 491/Brookhaven Medical Research Reactor, Brookhaven National Laboratory, 1997.

Medical Department Chairman, Review of Potential Environmental Release Points, Medical 

Department, Building 931, Brookhaven National Laboratory, 1997.

Chemistry Department Chairman, Review of Potential Environmental Release Points, Chemistry 

Department, Building 906, Brookhaven National Laboratory, 1997.

RHIC Project Head, Review of Potential Environmental Release Points, RHIC Project, Building 902 

Complex, Brookhaven National Laboratory, 1997.

Chemistry Department Chairman, NSLS Building 725 Final Facility Review Report, Chemistry 

Department, Building 901, Brookhaven National Laboratory, 1997.

M. Kelly and W. Thomlinson, Review of Potential Environmental Release Points, Brookhaven National 

Laboratory, June 17,1997.

Stephen V. Musolino, Review of Potential Environmental Release Points, RHIC Project, RHIC 

Complex, Brookhaven National Laboratory, June, 1997.

Medical Department Chairman, Review of Potential Environmental Release Points, Revision 2, 

Medical Department/Division, Building/Facility: Building 801, Brookhaven National Laboratory, 

July 25, 1997.

Department of Advanced Technology Chairman, Review of Potential Environmental Release Points, 

Department of Advanced Technology, Building 830, Brookhaven National Laboratory, June 

1997.

Chemistry Department of Chairman, Review of Potential Environmental Release Points, Chemistry 

Department, Building 555, Brookhaven National Laboratory, June 1997.

Physics Department of Chairman, Review of Potential Environmental Release Points, Physics 

Department, Building 510 Including Trailers, Brookhaven National Laboratory, September 

1997.

RHIC Project Head, Review of Potential Environmental Release Points, RHIC Project Buildings 620, 

820, 835, 836, Brookhaven National Laboratory, June 1997.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 48 of 106

Central Shop Division Manager, Review of Potential Environmental Release Points, Physics 

Department/Division Central Shops, Division Building/Facility 530, Including, Brookhaven 

National Laboratory, June 1997.

Instrument Division Head, Review of Significant Environmental Release Issues, Instrument Division, 

Building 535, Brookhaven National Laboratory, June 1997.

Safety & Environmental Protection Division Head, Review of Potential Environmental Release Points, 

Safety & Environmental Protection Division Bldg. 855, 860, 865 and 870, Including, 

Brookhaven National Laboratory, June 1997.

Safety & Environmental Protection Division Head, Review of Potential Environmental Release Points, 

Safety & Environmental Protection/Plant Engineering, Building 650-650T/Radiological Waste 

Decontamination & Hot Laundry Facility, Brookhaven National Laboratory, August 1997.

Manager of Plant Engineering, Review of Potential Environmental Release Points, Safety & 

Environmental Protection/Plant Engineering, Building 650-650T/Radiological Waste 

Decontamination & Hot Laundry Facility, Brookhaven National Laboratory, August 1997.

J. R. Naidu and L. L. Olmer, et al., 1981 Environmental Monitoring Report, BNL 51724/UC-11, Safety 

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L. E. Day and J. R. Naidu, 1982 Environmental Monitoring Report, BNL 51697, Safety & 

Environmental Protection Division, Brookhaven National Laboratory Associated Universities, 

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J. R. Naidu, B. A. Royce, and R. P. Miltenberger, Site Environmental Report for Calendar Year 1992, 

BNL 52411, Safety & Environmental Protection Division, Brookhaven National Laboratory 

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J. R. Naidu and B. A. Royce, Site Environmental Report for Calendar Year 1993, BNL 52432, Safety 

& Environmental Protection Division, Brookhaven National Laboratory Associated Universities, 

Inc., May 1994.

G. L. Schroeder et al., Site Environmental Report for Calendar Year 1996, BNL 52543, Safety & 

Environmental Protection Division, Brookhaven National Laboratory Associated Universities, 

Inc., January 1998.

R. J. Lee, J. R. Naidu et al., Site Environmental Report for Calendar Year 1997, Environmental, 

Safety and Health Services Division and Environmental Restoration Division, Brookhaven 

National Laboratory/Brookhaven Science Associates, February 1999.

R. J. Lee, J. R. Naidu et al., Site Environmental Report 2000, BNL-52626, Brookhaven Science 

Associates, LLC, September 2001.

General Reference: 

Crease, Robert P., Making Physics, A Biography of Brookhaven National Laboratory, 1946 -1972, 

University of Chicago Press, ISBN 0 226 12017, 1999.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 49 of 106

3.0

3.1 INTRODUCTION

OCCUPATIONAL MEDICAL DOSE

U.S. Atomic Energy Commission-funded work at Brookhaven National Laboratory (BNL) began in 

1948. BNL clinicians prescribed chest X-rays as part of the pre-employment and routine physical 

examination. These radiographs caused exposure of the lungs and other tissues of the body. 

Exposure came from the primary X-ray beam and from scattered and leakage radiation. 

Substantial information regarding the historical BNL occupational X-ray program has been located. 

The current medical director and two BNL X-ray technicians were interviewed and a few older X-ray 

films were reviewed to gain insight into the occupational medical program and collimation practices 

(Morris 2006a, b, c, and d). A peer-reviewed publication (Handloser and Love 1951) provided 

exposure rate data specific to the BNL radiographic equipment and techniques in use for diagnosis of 

radioactive material workers in 1950. A transcript of interviews conducted in preparation for an 

epidemiology study (Brodsky 1960) included discussions with BNL‘s first medical director regarding 

the early radiographic examination practices, equipment and exposure rates. 

3.2 PURPOSE

The purpose of this section is to determine the radiation dose associated with the occupational 

medical X-ray program conducted at the BNL.

3.3 SCOPE

The scope of this effort involved reviewing records, interviewing current and previous X-ray

technicians and reviewing X-ray films selected at random and dating back to 1948. Peer-reviewed 

publications provided exposure-rate data specific to the BNL radiographic equipment and practices in 

use for diagnosis of radioactive-material workers in 1950. Review of transcripts of interviews with 

BNL’s first medical director revealed the early radiographic examination practices, equipment and 

exposure rate. 

3.4 EXAMINATION FREQUENCY 

The energy employee medical chart provided by DOE may provide information on the number and 

kind of X-rays administered to the energy employee. 

Beginning in 1947 and continuing until at least 1964, pile (reactor) operators and possibly the hospital 

medical staff were subjected to an entry and an annual physical examination, including a PA chest Xray. All other workers, including visiting scientists employed for more than three months, were on a 

less frequent schedule. This group received an entry physical and a re-examination, including a PA 

chest X-ray, on a frequency averaging about one exam every 2.5 years (Brodsky 1964 pp. 38 – 39 

and p.72). It is assumed this examination frequency pattern continued through 1978. 

In 1978 a Presidential Recommendation to federal agencies advised to not perform routine or 

screening radiographic examinations, including chest X-rays, on patients for whom no prior clinical 

evaluation of the patient is made (Carter, 1978). This recommendation, at least in part, was 

incorporated into BNL practice and the average time between radiographic examinations increased. 

Starting with 1979 the re-examination cycle is assumed for all employees to be once every 7 years 

(Morris 2006b). Pre-employment radiographic examination is assumed to continue as a condition of 

employment.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 50 of 106

At some date in the 1980s (the exact year is unknown) the typical radiographic examination expanded 

to include both PA and LAT chest films (Morris 2006a). For expedience this is assumed to start in 

1979 coinciding with the extended re-examination frequency. 

Current (as of 2006) pre-employment X-ray examination practice is to offer PA and LAT chest X-ray to 

all employees and it appears that most new employees choose to have that examination. For new 

employees who test positive to tuberculosis, pre-employment PA and LAT chest X-ray are required. 

No lumbar spine films are done as a pre-employment requirement. The time between routine chest 

X-rays has been extended over the years and now occurs on a 7 to 10 year cycle (Morris 2006b and 

c). 

3.5 EQUIPMENT AND TECHNIQUES

The first BNL occupational X-ray was taken in October 1947 using a machine known as the “Powers 

unit” which may have been a portable machine designed for tuberculosis screening using 

photosensitive paper as the image receptor. No specific data are available about the technique, 

filtration or collimation. However, in a retrospective interview, the medical director stated each film 

caused a 4,000 mR exposure (Brodsky 1964, p. 36). It is assumed that the primary beam was not 

collimated. Medical radiographic tubes used in the 1940s typically had inherent filtration of 0.5 mm Al 

(ORAUT 2005, p. 9). Assuming the unit operated at a nominal tube potential of 70 kV, this 

corresponds to an HVL of approximately 0.8 mm Al (NCRP 1989, Table B.2).

By 1948 or 1949 the hospital facilities at BNL were built and a “photocross pocket unit” was installed. 

No specific data are available about the technique, filtration or collimation. However, the medical 

director stated some of the techniques used created a 1,000 mR exposure. It is assumed that the 

primary beam was not collimated.

The “photocross pocket unit” was soon replaced with another unit, not otherwise described. This new 

unit reduced the exposure associated with a 14 × 17 inch film to between 50 and 75 mR. It is 

assumed that the primary beam was not collimated.

By 1951 the occupational X-ray program was well established and data regarding the equipment and 

exposure had been published in the journal of Radiology (Handloser and Love 1951). That paper 

listed X-ray tube data for a photofluoroscopic unit, a fluoroscopic unit and a diagnostic film unit. 

However, it is concluded that only the diagnostic unit was used routinely for examinations required as 

a condition of employment (Brodsky). The 14 × 17 inch PA chest film technique used for employee 

screening was done using a Dynamax 25 tube head, style number 981625, serial number 12173, with 

0.5 mm Al filtration operating at 200 mA and 1/10 second at 72 inches. The nominal tube potential 

used in this technique was 72 kV and only the tube potential was adjusted to compensate for patient 

size (Handloser and Love 1951). The HVL of the X-ray beam is expected to be 0.8 mm Al (NCRP 

1989, Table B.2).

An interview (Morris 2006a) with a registered X-ray technician who worked at BNL from 1960 to 1993 

revealed that a well-maintained, collimated General Electric machine capable of fluoroscopy was in 

use for chest exposures in 1960. The unit was set up in Room 2 and used almost exclusively for 

chest X-ray (Morris 2006c). The equipment was equivalent to the new X-ray machines in service at 

nearby Smithtown General Hospital. Although no specific information is available regarding filtration, 

it is assumed that the unit contained 2.5 mm Al filtration as recommended by the NCRP at that time 

(ORAU 2005, p. 9). The technique used for chest film remained consistent from 1960 to 1993 (Morris 

2006a). For a PA chest film the nominal technique was 100 kVp, 200 mA, 1/60 s at 72 inches and the 

exposure was about 5 mR. The LAT chest film nominal technique was 100 kVp, 200 mA, 1/30 s at 72 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 51 of 106

inches. The GE machine was in service until 1991 when it was replaced by a Picker unit that is 

currently in service in room 1. 

An interview (Morris 2006b) with a registered X-ray technician who worked at BNL from 1993 to the 

present (March 2006) revealed that the equipment currently in service is a single phase Picker model 

BCX with a measured half value layer (HVL) of 3.2 mm Al. The unit is automatically collimated. A 

grid cassette film holder is routinely used. This unit has been in service since 1991. The unit is tested 

routinely by a medical physicist in accordance with New York State Department of Health Guidelines. 

Test records estimate tube filtration as 4.06 mm Al and the HVL as 3.3 mm Al (ASTARITA, 2004). 

For a PA chest film the nominal technique in current use is 110 kVp and 3.2 mAs at 72. No 

measurements were found for the LAT chest film technique but the technique is similar to PA except it 

is performed at 6 mAs. 

A sample of 26 PA chest films spanning the years 1949 through 1971 were reviewed to determine 

past beam collimation practices. The equipment in use appears to have allowed the technician to 

manually set the collimation. In some films prior to the early 1970’s collimation is evident. In other 

films from the same period no evidence of collimation is visible. That does not necessarily mean 

collimation was not used. It may indicate that collimation was established just at the edges of the film 

cassette and where therefore not visible on the film, which would be consistent with the practice 

described by the technician employed in that timeframe. The earliest film in this sample with obvious 

collimation was made in 1959 (Morris 2006b). Based on this inspection and the statement of the 

retired technician it is reasonable to assume that all exposures since 1960 were collimated.

3.6 ENTRANCE SKIN EXPOSURE (ESE)

In 1951 the BNL medical staff adopted the unusual (possibly unique) practice of recording the ESE for 

some examinations. Such information is found on the radiologist’s interpretation form “Report of X-ray

Examination” which is included in the employee’s medical chart. These ESE data were based on 

measurements made during the annual calibration of the X-ray machine. This practice of recording 

ESE was still occurring in 1964 (Brodsky 1964, p 67). These contemporary data reflect the exposure 

for a specific machine and a nominal, not patient-specific, technique. It is not known if this exposure 

included backscatter, and it does not appear to adjust for the patient’s chest thickness. So though it 

may be a more accurate assessment of the machine output at the time of the exposure it is not likely 

to significantly improve the accuracy of the dose reconstruction. 

1947

The first BNL medical director stated in a retrospective interview that each X-ray exposures made in 

1947 using the “Power[s] unit” caused a 4 R exposure (Brodsky 1964, p 36). 

1948 to 1949

In 1948 or 1949 the “photocross pocket unit” was in use creating a 1 R exposure. By 1950, the 

exposure associated with a 14 × 17 inch film was reduced to 50 to 75 mR (Brodsky 1964, p 36). This

is consistent with the specific information published in Handloser and Love (1951). 

1950 to 1960

Precisely measured exposure for diagnostic procedures was reported in a peer-reviewed journal 

(Handloser and Love 1951). PA chest films (14 × 17 inches) produced an exposure at the skin 

location, including backscatter, of approximately 54 mR. Exposure was measured with and without a 

paraffin phantom adjacent to the detector. The results were tabulated only for the measurements with 

the phantom, but they were displayed graphically for measurements with and without the phantom. 

Visual interpretation of the data shows the skin exposure was approximately 22% higher when the 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 52 of 106

phantom was in place. This is equivalent to a backscatter factor (BF) of 1.22. To determine ESE, 

which is exclusive of backscatter, the measured value with a phantom is divided by the BF resulting in 

an ESE for PA chest films of 44 mR. 

Although it seems unlikely that the following techniques were routinely used for examinations 

performed as a condition of employment, the following ESEs were also reported: 

• PA chest photofluoroscopy ranged from 700 to 1,200 mR, proportional to patient size, 

• AP lumbar spine 1,500 mR,

• LAT lumbar spine 5,700 mR.

1960 to 1990

By 1960 the GE equipment in Room 2 was in routine service for chest radiography (Morris 2006c). In 

1985 a measurement of the GE unit in Room 2 was performed at 100 kVp and 200 mA (Zukas 1987), 

the chest technique in use at the time (Morris 2006a). The exposure rate at the film for a 1/40 s 

exposure was 1.60 mR/mAs. When this is scaled to the 1/60 s exposure time used in the PA chest 

view and corrected based on the inverse squared relationship for a SSD of 154 cm the entrance skin

exposure for that view is calculated as 7.5 mR. When LAT view is scaled for exposure time, 1/30 s, 

and corrected for SSD the calculated entrance skin exposure is 15 mR. It is possible that higher tube 

potentials were used for the LAT chest view. For consistency with a rule of thumb (ORAU 2005, p 20) 

and to ensure this exposure is not underestimated it is rounded up to 20 mR.

1991 to present

Between 1996 and 2004 the PA chest ESE from the Picker unit in Room 1 was measured at least five 

times by ABR / ABMP certified radiation physics consultants during the course of routine inspection. 

The technique factors were similar to those used for nominal PA and LAT chest examination in 

current use, namely 110 kVp, 3.2 mAs on the large focal spot with a 72 inch SID and 23 cm phantom. 

The ESE ranged from 6.7 to 8.1 mR. For years 1991 to the present (2006) the ESE for PA chest is 

rounded up to 10 mR to account for the possibility that the actual technique used was set to a slightly 

higher tube potential. No measured ESE is available for the LAT view. Therefore the ESE for the 

LAT chest was calculated to be 25 mR based on the rule of thumb that the ESE for LAT chest is 

conservatively estimated to be 2.5 times the ESE for PA chest (ORAU 2005, p 20).

3.7 ORGAN DOSE CALCULATIONS

ICRP Publication 34 Method for collimated exposures

The methodology of ICRP Publication 34 (1982) is used to estimate the organ doses when the 

tabulated dose conversion factors (DCF) for collimated X-ray beams are applicable. The method is 

described in section 3.7.1, Organ Dose from Collimated Exposure. Modifications to this method to 

account for poorly collimated X-ray beams are described in section 3.7.2, Organ Dose from 

Uncollimated Exposure. Calculation of skin dose is described in section 3.7.3, Skin Dose.

Tables 3-1 and 3.2 are summaries of parameters, organ doses and skin doses for 14-in. by 17-in. PA 

and LAT chest radiographs respectively, for the periods during which each X-ray system was in use. 

In these tables the remainder organs are assigned the same dose as the female lung which is the 

organ with the highest DCF. 

Tables 3-3 and 3-4 list, by year, certain organ doses useful in a dose reconstruction for PA chest and 

LAT chest, respectively. When the organ dose differs between male and female, the larger of the two 

is listed. In these tables both the maximum additional organs and the maximum remainder organs are 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 53 of 106

assigned the same dose as the female lung which, for chest projections, is the organ with the highest 

DCF. 

3.7.1

In ICRP 34 (1982) methodology, organ dose (OD) is the product of entrance kerma (EK) and a dose 

conversion factor (DCF). Entrance kerma is defined in ICRP 34 (1982) as “air kerma in air without 

backscatter.” For practical application it is numerically the same as, and can be used interchangeable 

with, entrance skin exposure (ESE) (ORAUT 2005, p. 16). For organs other than skin, organ dose is 

calculated as shown in Eq. 3.1.

Organ Dose from Collimated Exposure

OD EK ×= DCF (3.1) 

DCF values are a function of beam quality, expressed as the HVL, and of the X-ray projection and 

organ of interest, assuming that the primary beam is appropriately collimated.

DCFs for some organs are not tabulated in ICRP 34 (1982) but are needed for the Interactive 

RadioEpidemiological Program (IREP). In those cases a dosimetric analogy to a nearby organ is 

assumed and the missing DCF is assigned the same value as the analogue organ. ICRP reference 

organs and their IREP organ analogues have been defined (ORAUT 2005, Table 3-2). For some 

organs ICRP 34 (1982) lists gender-based DCFs and in that situation the larger of the two values is 

assigned as the DCF to the IREP organ analogues. 

3.7.2

Organ doses from poorly collimated radiographs are calculated in a similar manner as for collimated 

exposures. However, in those cases where the DCFs are not available, or are not applicable, 

guidance is provided in ORAU (2005). Overriding any default guidance is the preference for actual 

measurement data when available (ORAU 2005, p. 15).

Organ Dose from Uncollimated Exposure

For exposures prior to 1960 the HVL is known or assumed to be less than 1 mm Al. ICRP 34 DCF

values are reported only for beam quality in which the HVLs range from 1.5 to 4 mm Al. To address 

these data shortfalls, DCFs tabulated for HVLs of 1.5 mm Al are used. If the DCFs for an HVL of 0.8 

mm Al are derived from a power function fitted to the ICRP 34 data, they would, on average, be lower 

by about 50%. 

Some organs, specifically the thyroid, ovaries, uterus and testes, are not in the primary beam of a 

properly collimated PA chest projection. But if the beam is poorly collimated those organs may be in 

the primary beam. Therefore a DCF from a different projection is substituted to avoid underestimation 

of the organ dose. ORAUT (2005) Table 9-11 defines appropriate substitutions. These substitutions 

are incorporated in the data shown in Tables 3.1 and 3.3.

3.7.3

Skin doses from AP and LAT chest projections are calculated differently depending on proximity to the 

primary beam. The primary beam entrance skin dose (ENSD), primary beam exit skin dose (EXSD), 

entrance skin dose outside but near the primary beam (ENSDNPB), exit skin dose outside but near 

the primary beam (EXSDNPB), and remote skin dose (RSD) can be calculated.

Skin dose

ENSD is the product of ESE and a backscatter factor (BF) as shown in Eq. 3.2. The BF is from Table 

B-8 of NCRP 102 (NCRP 1989) or from empirical data (Handloser and Love 1971). The ENSD 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 54 of 106

applies to all skin surfaces in the beam on the entrance side of the body. For collimated chest 

projections this includes the skin of the back, back of the neck, back of the shoulders, and upper lower 

back. For poorly collimated chest projections this includes the skin of all of the above, plus skin of the 

upper arms, elbows, and ears.

ENSD ×= BFESE (3.2) 

EXSD is the ENSD divided by a adjusted absorption factor (AFF) which is the absorption factor (AF) 

for the technique tabulated in Table B.7 of NCRP Report No. 102 (NCRP 1989) decreased by 10% as 

shown in equations 3.3 and 3.4 respectively. This dose applies to all skin surfaces in the beam on the 

exit side of the body. For collimated chest projections this includes the skin of the front of the neck,

the front of the shoulders, and upper abdomen. For poorly collimated chest projections this includes 

the skin of the all of the above, plus the face below the eyebrows, upper arms, elbows, and ears.

AFF EXSD = ENSD (3.3) 

AFF AF ×= 9.0 (3.4) 

ENSDNPB is 10% of the ENSD as shown in equation 3.5. This equation is based on the tabulated 

values in ICRP 34 showing that the dose to the testes is 10% of the dose in the primary beam when 

the testes are located just outside the primary beam.

ENSDNPB ENSD ×= 1.0 (3.5) 

Similarly EXSDNPB is 10% of the EXSD as shown in equation 3.6. 

EXSDNPB ×= 1.0 (3.6) 

RSD, shown in equation 3.7, is a function of:

• ENSD,

• square of the distance from the center of the primary beam to the location of interest on the 

skin, 

• the ratio of scattered to incident exposure based on the exposure at 1 m due to 90 degree 

scattering of 70 kVp radiation per Table B-2 of NCRP Report No. 49 (NCRP 1976),

• average depth dose (ADD) at 12 cm (mid chest depth) for the technique tabulated in NCRP 

102 Table B.8, and

• a bias factor of 1.1, which is favorable to the claimant, allows for 10% uncertainty in tabulated 

values of ADD.

2 1 .0 0005 1.1 



 





××××=

R

RSD ENSD ADD (3.7) 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 55 of 106

3.8 DOSE RECONSTRUCTION

Tables 3-1 through 3-4 may be used for screening calculations to estimate the number and type of Xrays a worker may have received and the organ dose from each one. The medical records provided 

by DOE may include adequate information to define the date, type, and count of X-ray examinations 

that were administered to the claimant as a condition of employment. Only radiographs made as a 

condition of employment are to be included in dose reconstructions. Such exposures are limited to 

PA and LAT chest films and were required or recommended as part of a pre-employment examination 

or a routine reexamination. 

If confusion about the radiographic exposure record exists, consider requesting that the notes on the 

exterior of the envelope(s) containing the worker’s X-ray films be transcribed and provided. These 

notes should give insight to the reason that the exposures were made, for example pre-employment 

examination, routine surveillance, or diagnosis of injury. Assume that any radiograph that was not a 

PA chest or LAT chest was diagnostic, not to be included in dose reconstruction. If the X-ray envelop 

notes associate the annotation “LMD” or “industrial” with a particular exposure, that means the 

radiograph was a diagnostic exposure (i.e., associated with a workplace injury). “LMD” radiographs 

are not included in dose reconstruction. 

Depending on equipment and technique, the photon energy associated with occupational medical Xray dose is in the 30-to-250-keV energy group. Assignment of all BNL occupational medical doses to 

the 30-to-250-keV energy group is favorable to the claimant and recommended. 

3.9 UNCERTAINTY

ORAU (2005) analyzed uncertainties in the occupational medical X-ray organ doses. The document 

considered several major sources of uncertainty: measurement errors; variations in applied voltage 

(peak voltage), beam current and exposure time; and uncertainties due to worker size and placement. 

ORAU (2005) assesses the relative error in an individual ESE or organ dose to be ±30% at 1 standard 

deviation. For adjustment that is favorable to the claimant, the actual doses could have been as much 

as 30% larger than those listed in Tables 3-1 through 3-4.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 56 of 106

Table 3-1. Parameters and organ doses in rem for 14-in. by 17-in. PA chest radiography. 

Period Frequency Applicability

PA chest 

ESE (mR)

Organ doses (rem) from 14-in by 17-in PA chest radiography

Thyroida,c Ovariesa,d Testesa,d Lungsa Breasta 

Uterus 

(embryo)a,d

Bone 

marrowa WBa Remainderb

1947 Entrance: yes

Routine reexamination: annual

Exit: no

Pile (reactor) 

operators and 

medical staff

4000

Assumed to 

be 

uncollimated

4.8E-01 2.8E-01 1.2E-02

Male

9.7E-01

Female

1.0E+00

7.2E-02 2.6E-01

Male

2.0E-01

Female

1.7E-01

Male

3.3E-01

Female

2.6E-01

1.0E+00

1947 Entrance: yes

Routine reexamination: 

approximately 2.5 year interval

Exit: no

All other 

workers 

1948 to 

1949

Entrance: yes

Routine reexamination: annual

Exit: no

Pile (reactor) 

operators and 

medical staff

1000

Assumed to 

be 

uncollimated

 

1.2E-01 6.9E-02 3.1E-03

Male

2.4E-01

Female

2.5E-01

1.8E-02 6.45E-02

Male

4.9E-02

Female

4.3E-02

Male

8.3E-02

Female

6.6E-02

2.5E-01

Entrance: yes

Routine reexamination: 

approximately 2.5 year interval

Exit: no

All other 

workers

1950-

1960

Entrance: yes

Routine: annual

Exit: no

Pile (reactor) 

operators and 

medical staff

44

Assumed to 

be 

uncollimated

5.3E-03 3.0E-03 1.4E-04

Male

1.1E-02 

Female

1.1E-02

7.9E-04 2.8E-03

Male

2.2E-03

Female

1.9E-03

Male

3.7E-03

Female

2.9E-03

1.1E-02

Entrance: yes

Routine reexamination: 

approximately 2.5 year intervals

Exit: no

All other 

workers

1960 to 

1978

Entrance: yes

Routine: annual 

Exit: no

Pile (reactor) 

operators and 

medical staff

7.5 2.4E-04 7.5E-06 7.5E-08

Male

3.1E-03

Female

3.4E-03

3.7E-04 9.8E-06

Male

6.9E-04

Female

6.5E-04

Male

9.8E-04

Female

8.9E-04

3.4E-03

Entrance: yes

Routine reexamination: 

approximately 2.5 year intervals

Exit: no

All other 

workers

1979 to 

1991

Entrance: yes

Routine: assumed to be at 7-yr 

intervals 

Exit: no

All workers

1991 to 

present

Entrance: yes

Routine: assumed to be at 7-yr 

intervals 

Exit: no

All workers

10 6.2E-04 3.2E-05 1.0E-07

Male

5.7E-03

Female

6.1E-03

9.1E-04 3.0E-05

Male

1.5E-03

Female

1.4E-03

Male

1.7E-03

Female

1.6E-03

6.1E-03

a. Organs identified in ICRP (1982) for dose determination from ESE associated with chest radiography.

b. Female lung is the reference organ for “Remainder”.

c. For uncollimated projections the DCF for thyroid is based on the AP cervical spine projection as showm in ORAUT (2005) Table 9-11. 

d. For uncollimated projections the DCF for ovaries, testes, and uterus is based on the PA abdominal projection as shown in ORAUT (2005) Table 9-11. 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 57 of 106

Table 3-1 (Continued). Parameters and organ doses in rem for 14-in. by 17-in. PA chest radiography. 

Period Frequency Applicability

Skin doses (rem) from 14-in by 17-in PA chest radiography

Entrance 

skin dose in 

primary 

beam

Exit skin 

dose in 

primary 

beam

Entrance skin 

dose near but 

outside 

primary beam

Exit skin 

dose near but 

outside 

primary beam

Remote 

skin dose 

– thighs to 

kneese

Remote 

Skin dose – 

knees to 

anklesf

Remote skin 

dose – 

anklesg

1947 Entrance: yes

Routine reexamination: annual

Exit: no

Pile (reactor) 

operators and 

medical staff

4.9E+00h 3.3E-02j 4.9E-01 3.3E-03 7.3E-04 2.7E-04 1.2E-04

1947 Entrance: yes

Routine reexamination: approximately 

2.5 year interval

Exit: no

All other 

workers

1948 to 

1949

Entrance: yes

Routine reexamination: annual

Exit: no

Pile (reactor) 

operators and 

medical staff

1.2E+00h 8.1E-03j 1.2E-01 8.1E-04 1.8E-04 6.5E-05 2.9E-05

Entrance: yes

Routine reexamination: approximately 

2.5 year interval

Exit: no

All other 

workers

1950-

1960

Entrance: yes

Routine: annual

Exit: no

Pile (reactor) 

operators and 

medical staff

5.4E-02h 3.6E-04j 5.4E-03 3.6E-05 8.0E-06 2.9E-06 1.3E-06

Entrance: yes

Routine reexamination: approximately 

2.5 year intervals

Exit: no

All other 

workers

1960 to 

1978

Entrance: yes

Routine: annual

Exit: no

Pile (reactor) 

operators and 

medical staff

1.0E-02i 2.3E-04k 1.0E-03 2.3E-05 3.3E-06 1.2E-06 5.4E-07

Entrance: yes

Routine reexamination: approximately 

2.5 year intervals

Exit: no

All other 

workers

1979 to 

1991

Entrance: yes

Routine: assumed to be at 7-yr 

intervals 

Exit: no

All workers

1991 to 

present

Entrance: yes

Routine: assumed to be at 7-yr 

intervals 

Exit: no

All workers

1.4E-02i 4.3E-04l 1.4E-03 4.3E-05 5.4E-06 2.0E-06 9.1E-07

e. Remote Skin Dose – Thighs to Knees is calculated at 0.52 m from the beam center line.

f. Remote Skin Dose - Knees to Ankles calculated at 0.86 m from the beam center line.

g. Remote Skin Dose – Ankles calculated at 1.28 m from the beam centerline. 

h. EK times BSF of 1.22 from Handloser and Love, 1951. 

i. Ek times BSF of 1.35 from NCRP 102. 

j. 1.5 mm Al HVL and chest thickness of 25 cm. 

k. 3.0 mm Al HVL and chest thickness of 25 cm. 

l. 4.0 mm Al HVL and chest thickness of 25 cm. 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 58 of 106

Table 3-2. Parameters and organ doses for 14-in. by 17-in. LAT chest radiography. 

Period Frequency Applicability

LAT 

chest 

ESE (mR) Thyroida Ovariesa Testesa Lungsa Breasta 

Uterus 

(embryo)a 

Bone 

marrowa WBa Remainderb 

Pre-1979 Entrance: none

Routine annual: none

Exit: none

All workers NA NA NA NA NA NA NA NA NA NA

1979 to 

1990

Entrance: likely

Routine assumed to be at 7 year intervals

Exit: none

All workers 20 2.3E-03 1.2E-05 2.0E-06 Male

3.9E-03

Female

4.4E-03

5.1E-03 1.2E-05 Male

7.4E-04

Female

5.8E-04

Male

1.2E-03

Female

1.3E-03

4.4E-03

1991 to 

present

Entrance: likely

Routine assumed to be at 7 year intervals

Exit: none

All workers 25 3.8E-03 4.0E-05 2.5E-06 Male

6.9E-03

Female

7.8E-03

7.9E-03 3.5E-05 Male

1.5E-03

Female

1.2E-03

Male 

2.4E-03

Female

2.2E-03

7.8E-03

a. Organs identified in ICRP (1982) for dose determination from ESE associated with chest radiography. 

b. Female lung is the reference organ for “Remainder”.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 59 of 106

Table 3-2 (Continued). Parameters and organ doses in rem for 14-in. by 17-in. LAT chest radiography. 

Period Frequency Applicability

Skin dose in rem from 14-in × 17-in LAT chest radiography

Entrance 

skin dose 

in primary 

beamc 

Exit skin 

dose in 

primary 

beamd

Entrance skin 

dose near but 

outside 

primary beam

Exit skin dose 

near but 

outside 

primary beam

Remote 

skin dose 

– thighs to 

kneesc

Remote 

skin dose – 

knees to 

anklesd

Remote skin 

dose – 

anklese

Pre-1979 Entrance: none

Routine annual: none

Exit: none

All workers N/A N/A N/A N/A N/A N/A N/A

1979 to 

1990

Entrance: likely

Routine assumed to be at 7 year 

intervals

Exit: none

All Workers 2.7E-02 1.3E-04 2.7E-03 1.3E-05 4.7E-06 1.7E-06 7.7E-07

1991 to 

present

Entrance: likely

Routine assumed to be at 7 year 

intervals

Exit: none

All workers 3.5E-02 2.3E-04 3.5E-03 2.3E-05 7.4E-06 2.7E-06 1.2E-06

 

c. Remote Skin Dose – Thighs to Knees is calculated at 0.52 m from the beam center ling. 

d. Remote Skin Dose – Knees to Ankles calculated at 0.86 m from the beam centerline. 

e. Remote Skin Dose – Ankles calculated at 1.28 m from the beam centerline.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 60 of 106

Table 3-3. Maximum organ doses (rem) for 14-in. by 17-in. PA 

chest radiography. 

Year

Maximum 

thyroid

Maximum 

ovaries

Maximum 

additional 

organsa Maximum remainderb 

1947 4.8E-01 2.8E-01 1.0E+00 1.0E+00

1948 to 1949 1.2E-01 6.9E-02 2.5E-01 2.5E-01

1950 to 1960 5.3E-03 3.0E-03 1.1E-02 1.1E-02

1961 to 1990 2.4E-04 7.5E-06 3.4E-03 3.4E-03

1991 to 2006 6.2E-04 3.2E-05 6.1E-03 6.1E-03

a. Not including skin.

b. Female lung is the reference organ for “Maximum additional organs” and 

“Maximum remainder”. 

Table 3-4. Maximum organ doses (rem) for LAT chest 

radiography. 

Year

Maximum 

thyroid

Maximum 

ovaries

Maximum 

additional 

organsa

Maximum 

remainderb 

1947 N/A N/A N/A N/A

1948 to 1949 N/A N/A N/A N/A

1950 to 1960 N/A N/A N/A N/A

1979 to 1990 2.3E-03 1.2E-05 4.4E-03 4.4E-03

1991 to 2006 3.8E-03 4.0E-05 7.8E-03 7.8E-03

a. Not including skin.

b. Female lung is the reference organ for “Maximum additional organs” and 

“Maximum remainder”. 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 61 of 106

REFERENCES

ASTARITA. Radiographic Skin Dose Charts 1998 – 2004. [SRDB REF ID:]

Brodsky, A. “Minutes of a meeting June 29, 1964”. [SRDB REF ID: 17344]

Carter, J, “Radiation Protection Guidance to Federal Agencies for Diagnostic X Rays,” in the Federal 

Register, Vol. 43, No. 22, February 1, 1978. [SRDB REF ID: 8141].

Handloser, J. S. and R. A. Love, 1951, “Radiation Doses from Diagnostic X-Ray Studies,” Radiology, 

volume 57, August, pp. 252–254. [SRDB REF ID:]

ICRP (International Commission on Radiological Protection), 1983, Protection of the Patient in 

Diagnostic Radiology, Publication 34, Pergamon Press, Oxford, England. [SRDB Ref ID 107]

Morris, R. L., 2006a, “BNL Occupational X-Ray: Interview with William Lehman, RT, X-Ray Technician 

(retired), Brookhaven National Laboratory,” documented communication, Oak Ridge 

Associated Universities Team, Oak Ridge, Tennessee, March 3. [SRDB REF ID:]

Morris, R. L., 2006b, “BNL Occupational X-Ray: Interview with Linda Cavaliere, RT, X-Ray 

Technician, Brookhaven National Laboratory,” documented communication, Oak Ridge 

Associated Universities Team, Oak Ridge, Tennessee, March 2. [SRDB REF ID:]

Morris, R. L., 2006c, “BNL Occupational X-Ray: Interview with William Lehman, RT, X-Ray Technician 

(retired), Brookhaven National Laboratory,” documented communication, Oak Ridge 

Associated Universities Team, Oak Ridge, Tennessee, March 28. [SRDB REF ID:]

Morris, R. L., 2006d, “email Dr Sbarra,” documented communication, Oak Ridge Associated 

Universities Team, Oak Ridge, Tennessee, March 2. [SRDB REF ID:]

NCRP (National Council on Radiation Protection and Measurements), 1989, Medical X-Ray, Electron 

Beam and Gamma-Ray Protection for Energies up to 50 MeV, Report No. 102, Bethesda, 

Maryland. [SRDB REF ID: 126]

ORAUT (Oak Ridge Associated Universities Team), 2005, Dose Reconstruction from Occupationally 

Related Diagnostic X-Ray Procedures, ORAUT-OTIB-0006, Rev. 03 PC-1, Oak Ridge 

Associated Universities, Cincinnati, Ohio, August 2. [SRDB REF ID: 6488]

Zukas, G., 1987, Brookhaven National Laboratory Memorandum to W. Lehman “X-ray Entrance 

Doses,” April 21. Subdocument in “X-ray Entrance Dose Surveys 83 to 96,” SRDB Ref ID 

project document number 030008980 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 62 of 106

4.0

4.1 INTRODUCTION

OCCUPATIONAL ENVIRONMENTAL DOSE

Ambient external gamma radiation outside radiologically controlled areas is the result of ubiquitous 

background radiation, or gamma emissions from stack effluent such as 41Ar, or skyshine due to air 

scatter from an otherwise well-shielded radiation source. 

During the early years (1950 to 1973), BNL used an ion chamber and dynamic capacitor electrometer 

assembly to measure ambient external radiation. The ion chamber/dynamic capacitor electrometer 

assemblies and thermoluminescent dosimeters (TLDs) were used concurrently during 1973 and 1974. 

The specific type of TLD used was not published until 1980 when the use of CaF2:Dy TLDs was 

noted (Naidu and Olmer 1981). The Site changed to LiF:Mg Tl technology in 2000 (BNL 2001). 

Monitoring results from up to four onsite stations are available for some years between 1967 and 

1984. When available, these data are included in the average dose estimates. Beginning in 1985, 

the perimeter and onsite monitoring program was expanded to approximately 20 onsite stations and 

has steadily increased to over 50 onsite stations as of 2004.

4.2 PURPOSE

The purpose of this section is to identify the sources and quantify the magnitude of the radiation dose 

received by BNL employees due to ambient radiation on the BNL site.

4.3 SCOPE

This section evaluates the ambient external radiation based on the monitoring results from four 

perimeter stations surrounding the BNL site at the northwest, southwest, southeast, and northeast 

coordinates. 

Doses due to inhalation and ingestion are evaluated using stack emissions and the results of BNL 

environmental reports.

4.4 AMBIENT EXTERNAL RADIATION

4.4.1

Prior to 1962, average and maximum environmental external gamma doses are derived from 

summary data published in retrospect (Meinhold and Meinhold 2001). During the period 1962 through 

1966, doses are derived from measurements at three or four continuous onsite monitoring stations as 

published in annual environmental monitoring reports. The standard deviation of the data for these 

years assumes a normal distribution. After 1966, on-site monitoring results were no longer reported 

and the only data available are fenceline measurements which in general are remote from areas 

occupied by employees. Consequently, after 1966 we conservatively assume that both the average 

and maximum annual doses are 100 mrem, the public dose limit. This assumed average value was 

exceeded only once in the prior years. The results, presented in Table 4-1, have been adjusted to 

reflect 2000 hour per year occupancy.

Use of Off-Site Monitoring
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 63 of 106

4.4.2

For purposes of dose reconstruction the ambient external dose should be applied to all unmonitored 

workers.

Applicability For Dose Reconstruction

Table 4-1. External gamma radiation dose (mrem/yr). 

Year

Monitoring 

location P-9

Monitoring 

location S-10

Monitoring 

location S-11

Monitoring 

location S-12

Site 

average

Error 1 

standard 

deviation Maximum

1947 0 0

1948 0 0

1949 0 0

1950 0.18 0.27

1951 6.4 9.2

1952 8.0 11

1953 8.5 12

1954 8.9 13

1955 8.9 13

1956 8.5 12

1957 7.3 11

1958 21 30

1959 27 39

1960 32 50

1961 27 39

1962 44 35 59 74 53 17 74

1963 135 141 125 127 132 8 141

1964 42 64 41 49 13 64

1965 40 97 38 58 34 97

1966 40 128 28 65 54 128

1967 to 

present

100 100

Notes: 

Monitoring locations P-9 was originally known as E-9.

Monitoring locations S-10, S-11 and S-12 were originally known as E-10, E-11 and E-12.

Data prior to 1950 are assumed based on site activities in those years.

Data from 1950 to 1961 are from Meinhold and Meinhold (2001).

Data from 1962 to 1965 are extracted from annual environmental monitoring reports

Data after 1965 are assumed, based on personnel dosimetry program criteria.

Error term includes only counting uncertainty and does not include sampling uncertainty.

4.5 INHALATION OF ONSITE AIRBORNE RADIONUCLIDES 

BNL has monitored releases to the environment since 1950 when the first stack became operational 

(Meinhold &Meinhold 2001). Stack release points and principal radionuclides in the airborne effluent 

are contained in the Site annual environmental reports.

4.5.1

Except for tritium, the intake for each radionuclide is the product of the highest 

Stack Monitoring

annual average concentration 

reported for that radionuclide at any stack

The individual stacks and their period of operation are:

or perimeter sampling location during the year and the assumed 

breathing rate of 3.3 × 10-4 m3

/s (ICRP 1975) based on a 2,000 hr/yr exposure.

Stack ID Period of operation Stack ID Period of operation

BGRR 1950–1969 Chem 1973 – 1993

HFBR 1966–2004 BMRR 1996 – 1997

Hot Lab & HFBR 1986 Incinerator 1981 – 1996
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 64 of 106

Hot Lab 1987–2004 Evaporator 1995 – 2000

BLIP/LINAC 1973–2004

The following isotopes were identified:

H-3 Co-60 Cs-137

Na-22 Zn-65 Ba-140

Na-24 Rh-106 La-140

Ar-41 Ru-106 Ce-141

Mn-54 I-125 Ce-144

Co-58 I-131 Eu-152

Fe-59 Cs-134 Eu-155

4.5.2

In Table 4-2 the intake of each radionuclide that substantially contributes to personnel dose is 

tabulated by year. As stated previously, except for tritium, the intake for each radionuclide is the 

product of the highest annual average concentration reported for that radionuclide at any stack or 

perimeter sampling location during the year and the assumed breathing rate of 3.3 × 10-4 m3

/s (ICRP 

1975) based on a 2,000 hr/yr exposure. The 3

H intake, assumed to be in the form of water vapor, is 

determined as described for other radionuclides then increased by a factor of 1.5 to account for direct 

absorption through the skin. 

Estimates of Potential Inhalation Intake

In most years stack effluent sampling shows that some radionuclides were released, but not 

specifically monitored at ground level site or perimeter sampling points. In those instances the 

highest annual average stack concentration was reduced by a factor of 0.01 to account for the 

lessened overall intake due to contribution from multiple, widely spaced facilities, atmospheric 

dispersion and building wake and the result was used as though it were a ground level site or 

perimeter sample. Most radionuclides released from the stacks were not subsequently detected by 

ground-level monitoring stations. Therefore, a reduction factor could not be calculated from the data, 

so a reduction factor of 0.01 was applied to the stack air concentrations to provide a bounding ground 

level concentration. 

No measurements of particle size or solubility are available. Solubility and particle size for 

environmental intake calculations should be assumed to be the same as those used in the internal 

dose section of this document. If no data are available default solubility classes and particle size 

values from the International Commission on Radiological Protection should be used (NIOSH 2002, 

pp. 15–16). 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 65 of 106

Table 4-2. Annual environmental occupational radionuclide inhalation (Bq/yr). 

Year H-3 I-131 1-125 Na-22 Na-24 Mn-54 Co-58 Fe-59 Co-60 Zn-65 Sr-90

Nb-95

Zr-95 Tc-99

1950 3.01E+04 2.70E-01 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1951 3.01E+04 4.73E+01 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1952 3.01E+04 6.09E+01 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1953 3.01E+04 6.42E+01 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1954 3.01E+04 6.76E+01 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1955 3.01E+04 6.76E+01 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1956 3.01E+04 6.42E+01 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1957 3.01E+04 5.41E+01 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1958 3.01E+04 6.15E+02 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1959 3.01E+04 6.15E+02 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1960 3.01E+04 6.15E+02 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1961 3.01E+04 6.15E+02 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1962 3.01E+04 7.03E+02 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1963 3.01E+04 6.15E+02 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 8.30E+02 N/A

1964 3.01E+04 5.71E+02 4.84E+04 5.51E-02 1.19E+03 3.52E+00 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 5.27E+00 N/A

1965 3.01E+04 7.45E+02 4.84E+04 5.51E-02 1.19E+03 6.15E-01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 5.27E+00 N/A

1966 3.01E+04 8.11E+02 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 8.79E-01 N/A

1967 2.11E+05 3.16E+02 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 4.84E+00 N/A

1968 1.71E+06 7.47E+05 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 4.84E+00 N/A

1969 1.71E+06 1.32E+01 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 2.64E-01 3.52E-01 6.68E+00 N/A

1970 3.96E+06 2.64E-01 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 7.91E-01 3.52E-01 8.79E+00 N/A

1971 1.05E+03 8.79E-02 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 3.52E-01 1.76E-01 8.09E+00 N/A

1972 7.91E+02 8.79E-02 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 8.79E-02 1.76E-01 1.58E+00 N/A

1973 3.41E+06 8.79E-02 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 8.79E-02 1.76E-01 2.64E-01 N/A

1974 4.46E+06 1.76E-02 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 1.26E+01 1.76E-01 3.08E+00 N/A

1975 3.42E+06 1.76E-02 4.84E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 1.22E+04 1.49E-03 8.79E-02 1.76E-01 1.05E+00 N/A

1976 4.08E+06 1.76E-02 4.84E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 1.22E+04 1.49E-03 8.79E-02 1.76E-01 1.05E+00 N/A

1977 4.11E+03 8.79E-01 4.84E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 1.22E+04 1.49E-03 8.79E-02 1.76E-01 8.88E+00 N/A

1978 5.32E+05 8.79E-01 4.84E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 1.22E+04 1.49E-03 8.79E-02 1.76E-01 8.88E+00 N/A

1979 1.37E+06 8.79E+00 4.84E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 1.22E+04 1.49E-03 3.42E+04 1.76E-01 8.88E+00 N/A

1980 2.05E+06 8.79E+00 4.84E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 1.22E+04 1.49E-03 3.42E+04 1.76E-01 8.88E+00 N/A

1981 1.32E+07 4.88E+03 4.84E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 1.22E+04 1.49E-03 3.42E+04 1.76E-01 8.88E+00 4.88E+03

1982 2.05E+07 4.88E+03 6.35E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 4.88E+03 1.49E-03 3.42E+04 1.76E-01 8.88E+00 4.88E+03

1983 1.90E+07 4.88E+03 8.30E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 4.88E+03 1.49E-03 3.42E+04 1.76E-01 3.52E-03 4.88E+03

1984 6.91E+06 1.47E+04 4.88E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 4.88E+03 2.37E-03 3.42E+04 1.76E-01 3.52E-03 4.88E+03

1985 9.66E+05 1.37E+00 2.98E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 1.47E+02 2.64E-02 1.40E-02 1.76E-01 3.52E-03 4.88E+03

1986 6.92E+06 1.03E+03 2.54E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 1.47E+02 1.15E+00 1.40E-02 1.76E-01 3.52E-03 4.88E+03

1987 1.15E+07 8.30E+03 4.35E+03 5.51E-02 1.19E+03 1.70E-03 3.22E-01 4.88E+01 1.15E+00 1.40E-02 1.76E-01 3.52E-03 2.05E+03

1988 1.71E+06 4.88E+02 9.52E+02 5.51E-02 1.19E+03 5.37E-01 3.22E-01 4.88E+01 4.36E-01 3.24E-01 1.76E-01 3.52E-03 2.44E-01

1989 2.70E+06 8.88E-02 1.03E+04 1.03E+00 1.19E+03 5.37E-01 3.22E-01 4.88E+01 6.44E-01 8.16E+00 1.76E-01 3.52E-03 2.44E-01

1990 4.63E+05 6.94E+02 2.30E+03 1.03E+00 1.19E+03 3.81E-02 3.22E-01 4.88E+01 6.55E+00 5.37E-03 1.76E-01 3.52E-03 2.44E+02

1991 6.44E+05 6.94E+02 4.75E+03 5.11E+00 1.19E+03 4.75E-02 1.34E+00 4.88E+01 2.83E-01 2.22E+00 1.76E-01 3.52E-03 2.44E+02

1992 4.44E+05 6.94E+02 2.59E+03 5.11E+00 1.19E+03 2.44E+01 1.34E+00 4.88E+01 2.77E+00 2.22E+00 1.76E-01 3.52E-03 2.44E+02

1993 5.33E+05 3.61E-02 2.59E+03 3.11E-01 1.19E+03 7.89E+00 6.83E+00 1.99E+01 6.65E-01 2.22E+00 1.76E-01 3.52E-03 2.44E+02

1994 5.92E+05 3.61E-02 2.44E+04 3.11E-01 1.19E+03 2.19E-01 6.83E+00 1.99E+01 9.49E-04 1.17E-01 1.76E-01 3.52E-03 2.44E+02

1995 7.15E+05 3.21E+00 1.95E+03 2.52E-03 1.19E+03 2.78E-02 6.74E-01 1.99E+01 3.84E-02 1.17E-01 1.76E-01 3.52E-03 2.44E+02
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 66 of 106

Table 4-2 (Continued). Annual environmental occupational radionuclide inhalation (Bq/yr).

Year H-3 I-131 1-125 Na-22 Na-24 Mn-54 Co-58 Fe-59 Co-60 Zn-65 Sr-90

Nb-95

Zr-95 Tc-99

1996 3.52E+05 3.21E+00 1.47E+03 1.12E-01 1.19E+03 4.30E-02 6.74E-01 1.99E+01 4.88E-03 1.01E+02 1.76E-01 3.52E-03 2.44E+02

1997 1.98E+05 1.72E+02 1.47E+03 4.55E-03 1.19E+03 7.42E-02 4.18E-01 1.99E+01 1.37E+01 1.01E+02 1.76E-01 3.52E-03 N/A

1998 2.73E+05 1.72E+02 1.47E+03 6.59E-03 1.19E+03 1.21E-02 1.99E-01 1.99E+01 7.03E-03 3.92E-01 1.76E-01 3.52E-03 N/A

1999 1.33E+05 1.72E+02 7.86E+00 6.59E-03 1.19E+03 9.33E-02 8.84E-02 1.99E+01 1.18E-01 2.17E+00 1.76E-01 3.52E-03 N/A

2000 3.52E+04 1.72E+02 7.86E+00 1.08E-03 1.19E+03 1.53E-02 1.45E-02 1.99E+01 1.93E-02 3.56E-01 1.76E-01 3.52E-03 N/A

2001 2.64E+04 3.83E-05 7.86E+00 1.08E-03 1.19E+03 1.53E-02 1.45E-02 1.99E+01 1.93E-02 3.56E-01 1.76E-01 3.52E-03 N/A

2002 2.37E+04 3.83E-05 7.86E+00 1.08E-03 1.19E+03 1.53E-02 1.45E-02 1.99E+01 1.93E-02 3.56E-01 1.76E-01 3.52E-03 N/A

2003 5.92E+05 3.83E-05 7.86E+00 1.08E-03 1.19E+03 1.53E-02 1.45E-02 1.99E+01 5.25E-08 3.56E-01 1.76E-01 3.52E-03 N/A

2004 5.92E+05 3.83E-05 7.86E+00 1.08E-03 1.19E+03 1.53E-02 1.45E-02 1.99E+01 5.25E-08 3.56E-01 1.76E-01 3.52E-03 N/A

Table 4-2 (Continued). Annual environmental occupational radionuclide inhalation (Bq/yr).

Year

Radionuclide

Tc-99m Ru-103

Rh-106

Ru-106 Cs-134 Cs-137 Ba-140 La-140 Ce-141 Ce-144 Eu-152 Eu-155 Ra-226 Th-228

1950 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1951 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1952 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1953 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1954 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1955 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1956 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1957 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1958 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1959 9.82E+03 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1960 9.82E+03 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1961 9.82E+03 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1962 9.82E+03 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1963 9.82E+03 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1964 9.82E+03 8.79E-01 1.23E+01 N/A 7.03E+00 1.23E+00 1.23E+00 8.79E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1965 9.82E+03 8.79E-01 2.20E+00 N/A 2.37E+00 1.23E+00 1.23E+00 8.79E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1966 9.82E+03 6.15E-01 5.27E-01 N/A 7.03E-01 1.23E+00 1.23E+00 4.40E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1967 9.82E+03 1.05E+00 5.27E-01 N/A 0.00E+00 1.23E+00 1.23E+00 4.40E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1968 9.82E+03 1.05E+00 5.27E-01 N/A 6.15E-01 5.27E-01 5.27E-01 4.40E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1969 9.82E+03 1.05E+00 2.11E+00 N/A 5.27E-01 1.76E-01 1.76E-01 4.40E-01 1.58E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1970 9.82E+03 1.05E+00 2.11E+00 N/A 1.23E+00 1.76E-01 1.76E-01 4.40E-01 2.46E+00 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1971 9.82E+03 1.05E+00 2.11E+00 N/A 1.05E+00 8.79E-02 8.79E-02 4.40E-01 6.15E-01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1972 9.82E+03 1.05E+00 2.11E+00 N/A 4.40E-01 2.64E-01 2.64E-01 4.40E-01 2.37E+00 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1973 9.82E+03 1.05E+00 1.23E+00 1.23E-01 8.79E-02 2.64E-01 2.64E-01 4.40E-01 2.37E+00 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1974 9.82E+03 1.05E+00 1.23E+00 1.23E-01 8.79E-02 2.64E-01 2.64E-01 4.40E-01 4.22E+00 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1975 9.82E+03 1.05E+00 6.15E-01 1.23E-01 8.79E-02 3.52E-01 2.64E-01 4.40E-01 6.68E+00 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1976 9.82E+03 4.88E+03 6.15E-01 1.23E-01 8.79E-02 3.52E-01 3.52E-01 4.40E-01 8.00E+00 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1977 9.82E+03 4.88E+03 7.91E-01 1.23E-01 8.79E-01 3.52E-01 3.52E-01 3.52E-02 1.32E+00 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1978 9.82E+03 4.88E+03 7.91E-01 1.23E-01 8.79E-01 3.52E-01 3.52E-01 3.52E-02 3.52E-01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1979 9.82E+03 4.88E+03 7.91E-01 1.23E-01 8.79E-01 2.20E-01 2.20E-01 3.52E-02 2.02E+00 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1980 9.82E+03 4.88E+03 7.91E-01 1.23E-01 8.79E-01 2.20E-01 2.20E-01 3.52E-02 7.91E-01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1981 9.82E+03 4.88E+03 7.91E-01 1.23E-01 3.96E-01 2.20E-01 2.20E-01 3.52E-02 7.91E-01 8.76E-07 6.07E-03 2.55E-01 1.76E-01
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 67 of 106

Table 4-2 (Continued). Annual environmental occupational radionuclide inhalation (Bq/yr).

Year

Radionuclide

Tc-99m Ru-103

Rh-106

Ru-106 Cs-134 Cs-137 Ba-140 La-140 Ce-141 Ce-144 Eu-152 Eu-155 Ra-226 Th-228

1982 9.82E+03 4.88E+03 9.67E-02 1.23E-01 2.64E+00 2.20E-01 2.20E-01 3.52E-02 5.36E-03 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1983 9.82E+03 4.88E+03 9.67E-02 1.23E-01 2.99E-01 2.20E-01 2.20E-01 3.52E-02 5.36E-03 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1984 9.82E+03 4.88E+03 9.67E-02 1.23E-01 1.77E-01 3.70E-03 2.20E-01 3.52E-02 5.36E-03 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1985 9.82E+03 5.86E+02 9.67E-02 1.23E-01 2.81E-01 3.70E-03 2.20E-01 3.52E-02 5.36E-03 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1986 9.82E+03 5.86E+02 7.03E-02 1.23E-01 2.81E-01 3.70E-03 2.20E-01 3.52E-02 5.36E-03 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1987 4.88E+02 5.86E+02 7.03E-02 9.58E-03 2.17E-01 3.70E-03 2.20E-01 3.52E-02 5.36E-03 8.76E-07 6.07E-03 2.55E-01 5.27E-02

1988 4.88E+02 1.12E+00 7.03E-02 9.58E-03 2.00E-01 3.70E-03 1.86E+02 3.52E-02 5.36E-03 8.76E-07 6.07E-03 8.44E-01 8.79E-02

1989 7.33E+03 1.12E+00 1.28E+02 2.60E-02 4.48E-01 3.70E-03 1.86E+02 3.52E-02 5.36E-03 8.76E-07 8.79E-03 6.15E-01 8.79E-02

1990 7.33E+03 1.52E+01 1.28E+02 2.60E-02 6.15E-01 3.70E-03 1.86E+02 3.52E-02 1.86E+00 8.76E-07 8.79E-03 1.03E+00 3.52E-02

1991 7.33E+03 1.52E+01 1.28E+02 2.60E-02 7.77E-01 3.70E-03 1.86E+02 3.52E-02 1.86E+00 8.76E-07 8.79E-03 1.86E-01 3.52E-02

1992 7.33E+03 1.52E+01 1.28E+02 2.60E-02 2.50E-01 7.76E+01 1.86E+02 9.41E-01 1.86E+00 8.76E-07 8.79E-03 1.86E-01 1.55E-01

1993 7.33E+03 1.52E+01 1.28E+02 2.60E-02 4.55E-01 7.76E+01 1.86E+02 9.41E-01 1.86E+00 8.76E-07 8.79E-03 1.86E-01 1.55E-01

1994 3.03E+02 1.52E+01 1.28E+02 4.49E-03 1.38E+00 7.76E+01 1.86E+02 9.41E-01 1.86E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

1995 3.03E+02 1.52E+01 1.28E+02 4.49E-03 2.76E+01 7.76E+01 1.86E+02 9.41E-01 1.86E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

1996 3.03E+02 1.52E+01 1.28E+02 4.49E-03 5.86E-01 7.76E+01 1.86E+02 9.41E-01 1.86E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

1997 3.03E+02 1.52E+01 1.28E+02 4.49E-03 1.47E-01 8.07E+02 4.27E+03 9.41E-01 7.34E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

1998 3.03E+02 1.52E+01 1.28E+02 4.49E-03 1.31E-02 8.07E+02 4.27E+03 9.41E-01 7.34E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

1999 3.03E+02 1.52E+01 1.28E+02 4.49E-03 2.59E-01 8.07E+02 4.27E+03 9.41E-01 7.34E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

2000 3.03E+02 1.52E+01 1.28E+02 4.49E-03 4.25E-02 8.07E+02 4.27E+03 9.41E-01 7.34E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

2001 3.03E+02 1.52E+01 1.28E+02 4.49E-03 4.25E-02 8.07E+02 4.27E+03 9.41E-01 7.34E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

2002 3.03E+02 1.52E+01 1.28E+02 4.49E-03 5.44E-04 8.07E+02 4.27E+03 9.41E-01 7.34E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

2003 3.03E+02 1.52E+01 1.28E+02 4.49E-03 5.44E-04 8.07E+02 4.27E+03 9.41E-01 7.34E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

2004 3.03E+02 1.52E+01 1.28E+02 4.49E-03 1.81E-09 8.07E+02 4.27E+03 9.41E-01 7.34E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

NOTES:

Highlighted values are bounded by available data. Missing years filled in w/ nearest results (highest value if uneven interval). 

Values not highlighted are from closest available result.

Tc-99 reported for incinerator only which operated from 1981-1996.

Tc-99m and Eu-155 are reported for BMRR which began operating in 1959; Tc-99m also reported for incinerator.

Cs-134 reported for BLIP which began operating in 1973; also reported for evaporator.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 68 of 106

4.6 INGESTION

Potable water, supplied from on-site wells, provides the most likely pathway for radionuclide ingestion

at BNL. Environmental reports include the radionuclide concentration in potable water. Most of the 

observed radionuclides are naturally occurring and consistent with concentrations found in offsite 

regional water supplies. 

Neither potable groundwater nor soil ingestion have been found to be pathways of exposure. No 

ingestion dose is indicated.

4.7 UNCERTAINTY

A high degree of uncertainty is associated with environmental monitoring. Uncertainty is the result of 

many factors listed here in approximately order of importance: atmospheric dispersion models used 

to describe dilution of radionuclide concentration from the stack to the receptor; location of ground 

level monitoring stations to adequately represent the unmonitored worker; efficiency of radioactive 

iodine sampling media prior to use of charcoal collectors; average water intake; occupancy factors for 

workers on the site; average breathing rate; measurement detection limits; and accuracy of 

measurement results. Additional factors could be listed. The error associated with each factor , but 

can be reasonably expected to range from more than a factor of 10 for atmospheric dilution, to about 

20% for accuracy of measurement. 

Although the historical uncertainty cannot be substantially reduced without a great deal of effort, 

importance of uncertainty in dose reconstruction is reduced by the fact that many of the doses are 

small. The largest annualized concentration or measurement result is used for intake or exposure, 

and other assumptions biased to be favorable to the claimant. An example is the dilution factor used 

to adjust stack effluent concentrations for use as ground level exposures. Actual data would suggest 

that roof level stacks tend to create dilution factors of 0.001 or more (ORAUT 2004, p. 28), but in this 

instance a factor of 0.01 is assumed which results in higher ground level exposure to an unmonitored 

worker.

The data presented provide an upper bound for annual external dose and radionuclide intake. The 

claimant-favorable biases incorporated into the analysis make a quantitative expression of uncertainty 

unnecessary. 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 69 of 106

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Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 72 of 106

5.0

5.1 INTRODUCTION

OCCUPATIONAL INTERNAL DOSE

BNL has played an important role in the development of the U.S. nuclear program. It has conducted 

applied research in nuclear and high-energy physics, chemistry and physics of materials, 

environmental and energy research, nonproliferation, neurosciences and medical imaging, and 

structural biology. Methods and concepts of measuring occupational internal doses to workers have 

evolved since the beginning of BNL operations. This section describes BNL internal dosimetry 

systems and practices.

5.2 PURPOSE

The purpose of this Section is to describe internal dosimetry systems and practices at the Brookhaven 

National Laboratory (BNL). This Section provides supporting technical data to evaluate internal 

occupational doses that can reasonably be associated with worker radiation exposures covered by 

the EEOICPA legislation.

5.3 SCOPE

This Section presents historical and current practices as they relate to the evaluation of internal 

exposure data for monitored and unmonitored workers. It describes plant facilities and processes, 

and historical information in relation to dose reconstruction for BNL workers. In addition it contains 

supporting documentation to assist in the evaluation of occupational internal doses from these 

processes in accordance with the Internal Dose Reconstruction Implementation Guideline (NIOSH 

2002) and addresses the evaluation of dose for unmonitored and monitored worker. 

The Section presents the technical basis of methods used to prepare dose information for input to the 

NIOSH IREP (Interactive RadioEpidemiological Program) computer code, and evaluates the 

uncertainty for BNL exposure and dose records.

5.4 THE BIOASSAY RECORDS IN THE INDIVIDUAL FILES 

5.4.1

The BNL bioassay record, Form 1720, contains one line (record) per sample. The record includes 

fields:

Brookhaven National Laboratory Bioassay Record (Form 1720)

1. NO. (number, i.e., the sequential number of the sample),

2. DATE RECV’D (date sample was received),

3. I or S (code, for incident or survey. Survey is assumed to be a routine sample), 

4. COLLECTION DATE/TIME,

5. COUNT DATE, 

6. TOTAL SAMPLE (volume in mL or X for a spot sample), and

7. SAMPLE TYPE (U for urine, F for feces, B for blood, H for hair, and WB for whole body).

The results portion of the form is filled in according to the analysis. Columns are provided for MFP 

(mixed fission products in dpm/day beta activity), 90Sr (in dpm/day beta activity), and 3

H (in μCi/L beta 

activity). Additional columns are provided for NUCLIDE, AMOUNT, and UNITS for alpha activity. 

Results for 90Sr only appeared with MFP results, indicating a sequential analysis. This form appears 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 73 of 106

to have been used from 1967 until at least 1986. Data from as far back as 1952 appears to have 

been transferred to these forms in 1967.

5.4.2

This form was used to calculate the whole body dose from tritium bioassay results. Results above 

some threshold (probably around 1 μCi/L) were transferred to this form. Each row was used to 

calculate the dose (if any) from one urine result. The column headings are:

Brookhaven National Laboratory Tritium Exposure Evaluation

1. No. of Intervening Days (since last sample), 

2. Gross μCi/L Current Date, 

3. Gross μCi/L Previous Date, 

4. Fraction Present of Previous Concentration,

5. μCi/L Remaining from Previous Date, 

6. Net μCi/L Current Date, 

7. Tritium Body Dose for Current Date MREM, 

8. Cumulative Tritium Body Dose MREM. 

Doses were not calculated if the Net μCi/L Current Date was less than some threshold (probably 

about 0.5 μCi/L). The form was used from the 4th quarter of 1968 until at least 1986. The doses were 

totaled for each quarter for comparison to the limits in place at the time. This form could be a 

secondary source of information should the Brookhaven National Laboratory Bioassay Record (Form 

1720) discussed above be missing or illegible. Based on the review of site records and personnel 

interviews, tritium dose was recorded on the external dosimetry cards (in the employee records) until 

the committed dose era. 

5.4.3

Summaries of the whole-body counting results may be included in some employees’ records for the 

1970s and possibly earlier. The summaries currently list several employees on a single page. Cover

letters included with the results provided an interpretation of positive results. For example, the letters 

discuss cesium results believed to be from fallout, results attributed to clothing contamination, results 

due to an intake with a body burden calculated, etc.

Summary of Whole-body Counting Results

5.4.4

By the 1980s, whole-body counting results for each count were included in the records. The earliest 

results were titled “NAI Results” and consisted of a one or two page computer printout. The counts 

appeared to be done with at single detector and were 15-minutes in length. The printout contains:

Individual Whole-body Counting Results

1. ACQUISITION DATE, date and time of the count 

2. ELAPSED LIVE TIME, length of the count 

3. NUCLIDE, a table of results listing the radionuclide 

4. CONCENTRATION (UNITS), activity 

5. ERROR, error.

By 1993, the site had changed to the Canberra ABACOS-plus software package. The ABACOS 

printout contained:

1. COUNT STARTED, the date and time of the count 

2. ELAPSED LIVE TIME, length of the count 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 74 of 106

3. NUCLIDE, a table of results listing the radionuclide 

4. ACTIVITY (nCi), the activity 

5. %ERROR (2 SD), and the error 

Initially the ABACOS printouts displayed “<” and the MDA in the results column for results below the 

MDA. Due to software limitations, an MDA and decision level could not both be displayed. In 

preparation for DOELAP accreditation, in 1999 the number displayed was changed to the decision 

level, but the printout incorrectly stated that the number displayed was the MDA (i.e., “MDA activity 

reported”). A list of MDAs was published separately. This was corrected when the software was 

upgraded and subsequently both a decision level and an MDA were reported.

5.5 IN VITRO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND 

REPORTING PROTOCOLS

5.5.1

In January 1949, a chemistry laboratory was organized under the Health and Safety Branch, Medical 

Division, New York Operations Office of the AEC to provide analytical and spectrographic facilities for 

dealing with hazards arising from NYOO facilities (e.g., uranium, beryllium, fluorine, thorium, and 

radium). In April, 1949, the NYOO chemistry laboratory began an investigation into the feasibility of 

using the levels of uranium excretion in urine as a control measure in production plants and 

processing facilities (NYOO 1949d). In 1953, this organization became the Health and Safety 

Laboratory (HASL) and in 1977 the Environmental Measurements Laboratory (EML). In 2002 it 

became part of the Department of Homeland Security. For convenience, this organization will be 

referred to as EML throughout this document.

In Vitro Urine Analysis

5.5.2

By 1951, BNL was doing its own research and development on urine sample preparation (BNL 1951). 

In 1952, an analysis procedure for fission products in urine was successful using an adaptation of the 

Chalk River method (BNL 1952). However, in 1955, some uranium in urine at BNL was still analyzed 

fluorometrically by EML (EML 1955). In 1958, a vibrating reed electrometer and chambers suitable for 

tritium urinalysis were set up in Building T-145 (Bishop 1958). In 1965, enriched uranium in urine was 

analyzed by EML. The analysis appears to have utilized both liquid scintillation counting (perhaps as 

a screening measurement) and electroplating for counting or autoradiography (Steimers 1965). The 

uranium fusion photofluorimetry urinalyses performed by the AEC’s New York Operations Office were 

similar to those performed at other AEC facilities. The default detection threshold for uranium 

urinalysis is assumed to be 10 μg/L based on a reported sensitivity of 5 to 10 μg/L for uranium 

fluorometric urinalysis in the early years (ORAUT 2005). During the period 1956 to 1964, the 

reporting level for uranium in urine by fluorometric analysis by EML appears to have been 1.0 μg/L 

(see for example EML 1957). Determination of 233U and 235U in urine was by fluorometry and alpha 

counting (EML 1956, 1961, 1962). It is uncertain whether the analyses by EML were for direct 

support, verification, or only for specific analyses needed. In May of 1963, 112 urine samples were 

processed by the in-house bioassay lab. Gross beta and 90Sr activity was measured in one individual 

who had been contaminated in April during reloading of spent fuel elements for shipment (BNL 1963).

History of In Vitro Urine Analysis

In 1973, bioassay analyses were performed by an Analytical Chemistry section (Bond 1973). In 1977, 

the Health Physics and Safety Division housed an analytical chemistry and bioassay laboratory 

(Liverman 1977). At least by 1999, Eberline was contacted for bioassay analyses, except for H-3 

which remained in-house until 2004 (~April 2004). Due to some problems with meeting the 

contractual MDAs, Eberline was replaced by the Severn-Trent (ST) Laboratory in December 2001. 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 75 of 106

ST also did the H-3 analyses when they were discontinued on site. The current BNL bioassay 

program is accredited by the DOE Laboratory Accreditation Program.

During the operation of the BGRR, some workers apparently were routinely sampled for mixed fission 

products and 90Sr. No program documentation was recovered but this seems to be the case from a 

limited review of the individual files. Sampling frequency varied from as short as monthly to as long 

as annually. After 1965 when the HFBR started operation, the most prevalent bioassay sampling was 

for tritium. During the operation of the HFBR, the concentrations of radionuclides other than tritium 

were not considered to be significant. The rationale was that concentration of tritium in the primary 

cooling water was several million times its “tolerance limit” where as fission products from fuel 

uranium surface contamination, such as 24Na, were present at less than 100 times tolerance.

During most of the site’s existence, it appears that department/division safety representatives or 

managers designated personnel for participation in the internal dosimetry program (WBCs and/or 

urinalysis) using guidance provided by Personnel Monitoring. This could also have included the 

frequency of monitoring (BNL 1995, Holeman 1999). Therefore, it may be difficult to determine if 

monitoring should have occurred by job title. For example, in the 1950s, several references 

mentioned that personnel described as “janitors” were involved in the clean-up of radioactive 

contamination. It was not clear if this was a specially designated crew or random site janitors being 

supervised by health physics personnel.

As some of the site’s older facilities have undergone decontamination and decommissioning (D&D) in 

recent years, D&D workers were sampled for 90Sr, 241Am, 238Pu, 239/240Pu, 234U, 235U, and 238U.

5.5.3

Although fecal sampling may have been used in response to specific events, no evidence of a 

comprehensive fecal sampling program was located.

Fecal Sample Analysis

Table 5-1 lists the frequencies for in vitro monitoring.

Table 5-1. Internal dose control program (in vitro). 

Routine monitoring type Period Frequency

Spot urine sample for H-3 No comprehensive documentation was located. It appeared to 

vary from weekly to monthly over time.

24-hour urine samples for fission, 

activation, uranium, and 

transuranics

No comprehensive documentation was located. In 1984, an 

annual program of urine sampling was in place for reactor 

workers. The analysis included 109Cd, 85Sr, 68Ga, 22Na, and 58Coa

a. Miltenberger and Steimers (1984). 

5.6 IN VIVO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND 

REPORTING PRACTICES

5.6.1

Whole body counting was established in 1960. It was first used as a research tool, but it became 

evident that it was also of use when someone was potentially involved in an incident. After confirmed 

intakes, whole body counting was repeated daily and the results were correlated with daily urine 

excretion results. Pile operators were the first group to be routinely counted. Some of the pile 

operators had been working before whole body counting started. Afterwards, all new pile operators 

received a whole body count before they started working and, starting in about 1962, annually 

thereafter. Other groups were apparently counted in the whole body counter as time permitted. 

Whole-Body Counting
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 76 of 106

Background counts were taken on unexposed individuals as a check on fallout radionuclides (e.g., 

cesium) in the general population (Love 1964). The counting facility, which is still in existence in the 

Medical Department, includes a counting bed with an array of sodium iodide thallium activated 

(NaI(Tl)) scintillation crystals.

By 1963 a portable “shadow shield” whole body counter was developed to measure the fission 

product body burdens of the Marshallese people who were exposed to the fallout from nuclear test 

Bravo on March 1, 1954 and to the subsequent uptake of radioactive material in crops. The detector 

was a 10 cm thick by 28 cm diameter NaI(Tl) crystal. The detector was stationary and was fixed in a 

position over the thorax of the subject. For over 25 years, BNL had a contract for the long-term 

medical surveillance of the Marshallese. In 1977, the responsibility for providing body burden 

measurements was transferred from the Medical Department to the Safety and Environmental 

Protection Division. In 1980, the minimum detection limit of the system was 1 nCi of 137Cs or 60Co for 

a 15-minute count (Miltenberger et al., 1980). In between these measurement trips, the counter was 

used at BNL as a part of the health and safety program and eventually performed the routine counts 

(Miltenberger 1981a). For positive results, additional counts could be done at the medical whole body 

counting facility to localize the contamination (Miltenberger 1981a). The minimum detection level 

(MDL) for the shadow shield arrangement was calculated in μCi by:

K

Ct MDL b

900 * 37 *

.4 65 =

Where:

Ctb = blank counts (background for region of interest)

900 = count time in seconds

37 = unit conversion factor

K = product of gamma abundance and counting efficiency for the radionuclide

By 1981, Cs-137 body burdens of 2 nCi or less were considered to be the result of fallout from 

atmospheric weapons testing environmental sources (Miltenberger 1981b).

In 1989, the MDA reported for thyroid counts for I-131 using a NaI counting system was “on the order 

of 10 nanocuries” (Miltenberger 1989).

The shadow shield replaced by stand up counter, but the date of this transition was not recovered. 

The NaI detectors were replaced by germanium detectors and eventually by broad energy germanium 

detectors.

In 2003, Canberra ABACOS 2000 software was used with the whole body counter. The MDA for 

each radionuclide was calculated by:

KT

S MDA .3 29 b + 3* =

Where:

Sb = sample standard deviation of background counts in the region of interest

K = efficiency at the centroid energy channel as determined arithmetically from the calibration 

efficiency equation times the yield abundance of the energy emission
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 77 of 106

T = count time

The ABACOS 2000 software produced a printout with the MDA and decision level for each result 

(Reciniello 2003).

5.6.2

Chest counts and thyroid counts were not done routinely. These counts were done as a result of 

incidents involving iodine (thyroid) and certain gamma emitters (127Xe, lung) (Miltenberger and 

Lessard 1987). Only one person who does iodinations was on a thyroid counting program in 2006.

Chest Counting

Table 5-2 lists the frequencies for in vivo monitoring.

Table 5-2. Internal dose control program (in vivo). 

Routine monitoring type Period Frequency

Pre-employment and annual 

WBCa

1962-unknown Annual for pile operators

Urine samples for H-3b 1993-unknown Weekly for reactor operators, supervisors, maintenance 

technicians, and S&EP building safety services technicians

WBCb 1993-unknown Start of service in Bldg 750, and termination of service for 

reactor division and S&EP personnel with offices in the HFBR 

building (750). Annual for reactor operators, supervisors, 

maintenance technicians, S&EP building safety services 

technicians and research coordination group technicians.

a. Love (1964)

b. Reciniello (1993)

5.7 UNCERTAINTY

At BNL the uncertainty for a single bioassay measurement was not reported consistently. 

Uncertainties associated with the analysis of MFP, 90Sr, or 3

H performed on site during the period 

from 1952 to 1986 were not included on the BNL Bioassay Record. From the 1980s to about 1998

uncertainties are reported on WBC sheets and are believed to be consistently reported as 2 sigma 

errors. Reviewed statements of work for bioassay services do not contain any specification for 

reporting uncertainty. Eberline reported a 1 sigma total error with its bioassay results. Currently, the 

Severn-Trent (ST) contract laboratory reports results with both a 1 sigma counting error and a 1 sigma 

total error.

5.8 DETECTION LIMITS

Table 5-3 lists the MDAs and reporting levels for periods corresponding to the bioassay methods 

discussed in Sections 5.3 and 5.4. The reporting levels are listed in the units quoted in the 

references, which are generally the units of the results. Detection limits are to be taken from the 

individual reports when available.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 78 of 106

Table 5-3. Detection limits.

Radionuclide Method/

description

Period MDAa Reporting levelb

Na-22 (1274.54 keV) WBC 1999 3 nCi

Mn-54 (834.83 keV) WBC 1999 3 nCi

Co-57 (122.06 keV) WBC 1999 7 nCi

Co-58 (810.76 keV) WBC 1999 4 nCi

Fe-59 (1099.22 keV) WBC 1999 4 nCi

Co-60 (1332.49 keV) WBC 1999 5 nCi

Zn-65 (1115.52 keV) WBC 1999 5 nCi

Cs-134 (604.7 keV) WBC 1999 3 nCi

Cs-137 (661.65 keV) WBC 1999 4 nCi

a. Values from Murray (1999). 

b. Reporting levels shown on lung count results reports are from ABACOS-Plus software. 

Table 5-3. Detection limits (continued) 

Radionuclide

Method/

description Period MDAa Reporting level

Be-7 (477.59 keV) WBC 2003 21.96, 22.2 nCi

Na-22 (1274.54 keV) WBC 2003 2.58, 3.21 nCi

K-40 (1460.81 keV) WBC 2003 36.50, 131.54 nCi

Mn-54 (834.83 keV) WBC 2003 2.22, 3.74 nCi

Co-57 (122.06 keV) WBC 2003 3.37, 2.44 nCi

Co-58 (810.76 keV) WBC 2003 2.49, 3.23 nCi

Fe-59 (1099.22 keV) WBC 2003 4.04, 6.07 nCi

Co-60 (1332.49 keV) WBC 2003 1.83, 3.65 nCi

Zn-65 (1115.52 keV) WBC 2003 5.18, 6.45 nCi

Sr-85 (513.99 keV) WBC 2003 2.38, 5.14 nCi

I-129 (29.62 keV) WBC 2003 7.97, 3.01 nCi

I-131 (364.48 keV) WBC 2003 2.51, 3.47 nCi

Cs-134 (604.7 keV) WBC 2003 2.09, 3.08 nCi

Cs-137 (661.65 keV) WBC 2003 2.64, 4.77 nCi

Ce-139 (165.85 keV) WBC 2003 3.48, 2.93 nCi

Ce-144 (133.54 keV) WBC 2003 27.12, 22.11 nCi

Bi-214 (609.31 keV) WBC 2003 7.00, 9.65 nCi

Pb-214 (351.92 keV) WBC 2003 7.02, 8.5 nCi

Am-241 (59.54 keV) WBC 2003 9.06, 4.94 nCi

a. Values for 2003 from Michel (2003b, 2003b) published for each calibration. 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 79 of 106

Table 5-3. Detection limits (continued)

Radionuclide

Method/

description Period MDAa Reporting level

Be-7 (477.59 keV) WBC 2004 23.35 nCi

Na-22 (1274.54 keV) WBC 2004 3.97 nCi

K-40 (1460.81 keV) WBC 2004 147.04 nCi

Mn-54 (834.83 keV) WBC 2004 2.50 nCi

Co-57 (122.06 keV) WBC 2004 1.40 nCi

Co-58 (810.76 keV) WBC 2004 2.39 nCi

Fe-59 (1099.22 keV) WBC 2004 6.10 nCi

Co-60 (1332.49 keV) WBC 2004 4.12 nCi

Zn-65 (1115.52 keV) WBC 2004 5.09 nCi

Sr-85 (513.99 keV) WBC 2004 2.21 nCi

I-129 (29.62 keV) WBC 2004 78.83 nCi

I-131 (364.48 keV) WBC 2004 2.20 nCi

Cs-134 (604.7 keV) WBC 2004 2.29 nCi

Cs-137 (661.65 keV) WBC 2004 3.62 nCi

Ce-139 (165.85 keV) WBC 2004 1.94 nCi

Ce-144 (133.54 keV) WBC 2004 13.59 nCi

Bi-214 (609.31 keV) WBC 2004 11.69 nCi

Pb-214 (351.92 keV) WBC 2004 8.32 nCi

Am-241 (59.54 keV) WBC 2004 5.73 nCi

a. Values for 2004 from Michel (2004) published for each calibration. 

Table 5-3. Detection limits (continued)

Radionuclide

Method/

description Period MDAa Reporting level

Be-7 (477.59 keV) WBC 2005 20.13 nCi

Na-22 (1274.54 keV) WBC 2005 2.76 nCi

K-40 (1460.81 keV) WBC 2005 42.99 nCi

Mn-54 (834.83 keV) WBC 2005 2.79 nCi

Co-57 (122.06 keV) WBC 2005 1.91 nCi

Co-58 (810.76 keV) WBC 2005 2.91 nCi

Fe-59 (1099.22 keV) WBC 2005 4.92 nCi

Co-60 (1332.49 keV) WBC 2005 2.90 nCi

Zn-65 (1115.52 keV) WBC 2005 5.54 nCi

Sr-85 (513.99 keV) WBC 2005 3.41 nCi

I-129 (29.62 keV) WBC 2005 81.99 nCi

I-131 (364.48 keV) WBC 2005 1.92 nCi

Cs-134 (604.7 keV) WBC 2005 2.89 nCi

Cs-137 (661.65 keV) WBC 2005 3.12 nCi

Ce-139 (165.85 keV) WBC 2005 1.76 nCi

Ce-144 (133.54 keV) WBC 2005 14.63 nCi

Bi-214 (609.31 keV) WBC 2005 7.72 nCi

Pb-214 (351.92 keV) WBC 2005 7.60 nCi

Am-241 (59.54 keV) WBC 2005 4.42 nCi

a. Values for 2005 from Michel (2005) published for each calibration. 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 80 of 106

Table 5-3. Detection limits (continued)

Radionuclide

Method/

description Period MDAa Reporting levelb

Am-241 (Eberline) Urinalysis 1999-2000 6.0 × 10-2 pCi/L No threshold

Am-241 Urinalysis 2001-2006 8.4 × 10-8 μCi/24 hours No threshold

Br-77 Urinalysis 1975 5 μCi/Le

H-3 (in house lab)c Urinalysis 1965-1979 2.0 × 10-2 μCi/L

H-3 (in house lab) Urinalysis 1996 4.0 × 10-3 μCi/L

H-3 (in house lab) Urinalysis 1999 3.3 × 10-3 μCi/L

H-3 (in house lab)d Urinalysis 2000 5.0 × 10-3 μCi/L

H-3 (required by TBD) Urinalysis 2001-2006 1.0 × 10-2 μCi/L

Mixed Fission Productsc Urinalysis 1952-1969 5.0 dpm/24 hours

Pu-238 (Eberline) Urinalysis 1999-2000 6.0 × 10-2 pCi/L No threshold

Pu-238 Urinalysis 2004-2006 8.4 × 10-8 μCi/24 hours No threshold

Pu-239 Urinalysis 2001-2006 8.4 × 10-8 μCi/24 hours No threshold

Pu-239/240 (Eberline) Urinalysis 1999-2000 6.0 × 10-2 pCi/L No threshold

Sr-90c Urinalysis 1950-1998 5.0 dpm/24 hours

Sr-90 (Eberline) Urinalysis 1999-2000 1.0 pCi/L No threshold

Sr-90 (required by TBD) Urinalysis 2001-2006 3.54 × 10-2 μCi/24 hours No threshold

Uranium (fluorometric) Urinalysis Up to 1955 10 μg/liter

Uranium (fluorometric) Urinalysis 1956-1965 1 μg/liter

U-238 Urinalysis 2001-2006 8.4 × 10-8 μCi/24 hours No threshold

U-238 (Eberline) Urinalysis 1999-2000 1.0 × 10-1 pCi/L No threshold

U-235 Urinalysis 2001-2006 8.4 × 10-8 μCi/24 hours No threshold

U-235 (Eberline) Urinalysis 1999-2000 1.0 × 10-1 pCi/L No threshold

U-234 (Eberline) Urinalysis 1999-2000 1.0 × 10-1 pCi/L No threshold

a. Values from Recinello et al. (2001) and Sun et al. (2004), except as noted. Values could be applicable earlier, but 

documentation was not located. 

b. Reporting levels shown could have been the MDA, but were not documented as such. 

c. Values from a reactor employee’s bioassay data. 

d. Value from Scarpitta (2001). 

e. Value below which “no exposure was assigned” from O’Connell (1975). 

5.9 EXCRETA SAMPLE KIT CODES

No codes have been found. Table 5-4 lists sample kit information summarized from various site 

documents.

Table 5-4. Excreta sample kit codes.

Kit codea

D/R P/U Media Sample description

NA NA Urine Routine samples for H-3 were collected as grab samples (1993) to be collected 

weekly or within 4 hours of completion of a task which required bioassay. Twentyfour-hour samples could also be collected when other radionuclides were 

suspected.a

 In 1984, 24-hour kits consisted of three 500 mL bottles.b In 1999, spot 

samples were collected in 60 cc plastic bottles.c

a. Reciniello (1993)

b. Miltenberger and Steimers (1984)

c. Holeman (1999)
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 81 of 106

5.10 SOLUBILITY TYPE, FRACTION ACTIVITY, AND PARTICLE SIZE BY FACILITY

In the absence of any measurements or studies, NIOSH guidance requires the use of default solubility 

classes and particle size values from the International Commission on Radiological Protection 

(NIOSH 2002, pp. 15, 16). Facility-specific solubility and particle size data for BNL has not been 

found. Activity fractions for occupational exposure were generally not available. For some facilities 

stack emissions or other measurements could be useful for dose reconstructions. Table 5-5 lists this 

information.

Table 5-5. Solubility type, fraction activity, and particle size by facility.

Facilitya Compoundb RN

Solubilityc

type

Particle sized

(μ AMAD) Activity fraction

Sewage 

Treatment Plant

(1962)

Unknown. Measured at the outfall from the 

Imhoff tank (Loysen 1962)

Sr-90 0.20

Cs-137 0.80

Sewage 

Treatment Plant

(1973)

Unknown. Fractions based on activities 

measured at the input to the sand filter beds 

(Liverman 1977, p. 126)

H-3 9.87E-01

Be-7 7.33E-03

Na-22 6.01E-05

Na-24 3.59E-03

Cr-51 2.49E-04

Co-58 9.02E-05

Co-60 2.58E-05

Zn-65 3.44E-05

Sr-90 5.58E-05

I-131 1.37E-03

Cs-134 1.72E-05

Cs-137 1.07E-04

Ce-144 1.98E-04

Brookhaven 

Graphite 

Research 

Reactor (BGRR)

Bldgs. 701-703, 

704 (fan house), 

708 (instrument 

house), 709 

(canal house), 

709A (canal 

water treatment 

facility)

(1950-1957)

Natural uranium metal fuel, 1 by 4 inch 

cylinders in 11 ft finned aluminum cartridges.

(Activity fractions based on estimates of core 

activity at the time of 28 fuel ruptures and one 

irradiated uranium sample failure) (Meinhold 

and Meinhold, 2001)

Ba-140 1.70E-01

Ce-144 1.08E-01

Cs-137 7.19E-03

La-140 1.95E-01

Nb-95 1.62E-01

Ru-103 7.28E-02

Ru-106 6.09E-03

Sr-89 1.17E-01

Sr-90 7.78E-03

Zr-95 1.53E-01

U-235 8.35E-08

U-238 1.81E-06

U-234 1.81E-06

Pu-239 2.86E-04
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 82 of 106

Table 5-5 (Continued). Solubility type, fraction activity, and particle size by facility.

Facilitya Compoundb RN

Solubilityc

type

Particle sized

(μ AMAD) Activity fraction

Brookhaven 

Graphite 

Research 

Reactor (BGRR)

Bldgs. 701-703

(1958-1969)

Enriched uranium fuel, in the form of curved 

plates of uranium-aluminum alloy, clad on all 

surfaces by 0.5 mm of aluminum. Routine 

releases of fission products occurred due to 

uranium contamination on the fuel’s surface 

and trapped in the cladding. Based on I-131 

and gross beta measurements in the 1967-

1969 EMRs.

Co-60 8.05E-02

Zr-95 3.79E-01

Ru-103 2.30E-01

Ce-141 1.26E-01

Ce-144 1.15E-01

Cs-137 6.90E-02

I-131 2.86E-03

Brookhaven 

Graphite 

Research 

Reactor (BGRR)

Bldgs. 701-703

(1970-2006)

Residual contamination resulting from fission, 

activation, and transuranic radionuclides 

produced in the core and spread primarily by 

contact with the cooling water (Musolino 

2000). Fuel storage canal was 

decommissioned in 1973.

Pu-238 6.66E-06

Pu-239/40 5.86E-04

Pu-241 2.59E-04

Am-241 5.05E-05

Fe-55 9.79E-01

Co-60 4.89E-04

Sr-90 9.09E-03

Cs-137 1.08E-02

U-235 1.37E-07

U-238 3.57E-06

Brookhaven 

Medical 

Research 

Reactor (BMRR)

Bldgs. 490-491

(1959-2000)

Enriched uranium fuel, in the form of curved 

plates of uranium-aluminum alloy, containing 

12% by weight of fully enriched uranium. 

Estimates are for 1997 routine releases of 

isotopes likely to be of dosimetric importance 

(Table 5, p 57, Meinhold and Meinhold, 2001) 

99.9% of activity release to sewer from labs 

was H-3 (Liverman 1977, p. 123)

Ce-141 2.99E-03

Ce-144 2.33E-02

Co-60 4.35E-02

Fe-59 6.32E-02

I-131 5.47E-01

Zn-65 3.20E-01

Cosmotron

Bldg.

(1953-1966)

Activation of short-lived isotopes in the 

ventilating air within the tunnel and 

experimental areas. See ”Accelerators & 

RHIC” below.

C-11

N-13

O-15

Alternating 

Gradient 

Synchrotron 

(AGS)

Bldgs. 905-912

(1960-2006)

Activation of short-lived isotopes in the 

ventilating air within the tunnel and 

experimental areas. Processing of irradiated 

targets (Meinhold and Meinhold, 2001, p. 58; 

Liverman 1977, p. 310). See ”Accelerators & 

RHIC” below.

C-11

N-13

O-15

AGS Bubble 

Chamber

Bldg. 919

Handling of tritium sources up to 100 mCi 

(1977)

H-3 1

Accelerators & 

RHIC

1953-2006

Posted contamination areas primarily in the 

vicinity of fixed targets (Lessard et al. 2001) 

Based on recent information, but is assumed 

to apply to earlier routine operations. Before 

1987, beam intensities were less but some 

uncontained targets (e.g., tungsten) were 

used.

Be-7 5.50E-01

Na-22 1.40E-01

Mn-54 1.40E-01

Co-57 1.40E-01

Co-60 3.00E-02

Hot Laboratory

(central lab, fan 

house, liquid 

waste tank 

farm, liquid 

waste concentration plant)

Bldg. 801, 802

(1951- 

Acid dissolution of irradiated uranium 

samples to recover radioiodines (March 1952 

to June 1960)

In 1957 there was an accidental release of 

unirradiated natural uranium hexafluoride due 

to an explosion (Meinhold and Meinhold, 

2001, p 59). 

Radioactive solutions up to mCi amounts, 

200 Ci/yr H-3 fabricated into accelerator 

targets (1977). 87% of the liquid waste in 

1973 was H-3 (Liverman 1977, p. 125)

Fractions based on environmental releases of 

isotopes of dosimetric importance, 1983-1991

Co-58 9.33E-02

Co-60 1.30E-03

Cs-137 1.92E-02

Eu-155 3.37E-03

I-131 8.67E-02

Mn-54 1.88E-03

Zn-65 7.91E-01

Zr-95 2.95E-03

U unknown

H-3 unknown

Physics

Bldg. 510

Processing irradiated targets, up to 1 mCi; 

5000 Ci Co-60 in the gamma irradiation 

facility (1977)
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 83 of 106

Table 5-5 (Continued). Solubility type, fraction activity, and particle size by facility.

Facilitya Compoundb RN

Solubilityc

type

Particle sized

(μ AMAD) Activity fraction

Radiation and 

Chemical 

Technology

Bldgs. 526-527

Unknown uranium of various enrichments. 

Originally housed a criticality facility for 

reactor physics (Liverman 1977, p 48). 

Monomer processing (Liverman 1977). 

Concrete dried and impregnated with various 

monomers.

High Intensity 

Radiation 

Development 

Laboratory

Bldg. 528

(1963-1972)

Contained million-curie Co-60 and Cs-137 

sources. The HIRDL was used for the 

development and testing of these sources. 

Operations were curtailed in 1972 due to lack 

of funding. In 1977, only a water-tank Co-60 

gamma irradiation facility was still in 

operation.

Cs-137

Co-60

Radiation 

Development 

Laboratory

Bldg. 830

Fabrication of solid sources, up to Ci 

amounts, monomer processing (1977)

Hot Machine 

Shop

Bldg. 530 

-1975

Replaced by

Bldg. 462

Machining radioactive solids (1977), including 

uranium and radioactive graphite; repair work 

on “hot” fuel handling tools.

Health Physics 

& Safety, 

Instrumentation

Bldg. 535

Radioactive solutions up to 100 μCi (1977)

Chemistry

Bldg. 555

Radioactive solutions and irradiated targets, 

up to 1 mCi (1977)

Chemistry 

Linac 

Irradiation 

Facility (CLIF)

Bldg. 931A

Production of radioactive isotopes. See 

”Accelerators & RHIC” above. Dates of 

operation were not located.

Reclamation 

and Hot 

Laundry

Bldg. 650

Radioactive decontamination, up to mCi 

amounts (1977)

82% of activity released to sewer in 1973 was 

H-3, rest was gross beta (Liverman 1977, p.

125)

High Flux Beam 

Reactor

Bldg. 750

(1965-1997)

“Fully” enriched uranium and aluminum alloy, 

heavy water cooled and moderated reactor 

that operated up to 40 megawatts (thermal) 

(60 MW planned in 1977)

Fractions based on activity in heavy water 

when systems were opened for refueling or 

maintenance (1977).

H-3 9.998E-01

Na-24 2.22E-04

Cr-51 2.22E-06

Co-59 8.89E-10

Co-60 6.67E-10

Zn-69 8.89E-09

I-131 6.67E-10

Cs-137 3.33E-10

Tandem 

Van de Graaff 

Accelerator

Bldg. 901, 901A

Tritium gas and vapor. Fractions based on 

activity released from facility in 1979 (Naidu 

1979)

H-3 Gas 8.95E-01

H-3 Vapor 1.05E-01

Superconducting Test, 

Machine Shop, 

RARAF

Bldg. 902

Irradiated target processing, up to 100 μCi. 

(1977)
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 84 of 106

Table 5-5 (Continued). Solubility type, fraction activity, and particle size by facility.

Facilitya Compoundb RN

Solubilityc

type

Particle sized

(μ AMAD) Activity fraction

Brookhaven 

Linac Isotopes 

Production 

Facility

(BLIP)

Bldgs. 930 & 

931B

Processing irradiated samples containing 

short-lived radionuclides (1977) for medical 

use; targets were sealed during normal 

operations but radioactive gases (N-13, O-14, 

O-15, N-16) and other isotopes formed in 

cooling water and equipment (Liverman 1977, 

p. 113). Fractions estimated based on 

activity in cooling and shield tank after one 

year of operation (Liverman 1977, p. 123).

H-3 HTO 6.35E-01

Be-7 3.05E-01

Na-22 2.54E-03

Na-24 5.08E-02

Cr-51 4.07E-03

Co-58 1.52E-03

Co-60 1.02E-03

Waste 

Management 

Incinerator

Bldg. 445

(1981-) 

BNL granted a permit by NY in March 1981. 

Fractions based on averages 1981-1991 for 

isotopes of importance for internal dose.

Fe-59 9.01E-02

I-131 1.03E-01

Mn-54 1.02E-02

Ru-103 2.87E-02

Ru-106 2.68E-03

Tc-99 5.01E-02

Zn-65 7.15E-01

a. Facilities were combined for this analysis if they were similar and had a common list of radionuclides of concern.

b. If chemical compounds were not available, the best description found is listed.

c. Specific information was not located, the solubility that is most favorable to the claimant should be applied. 

d. Specific information was not located, the 5 μm default is assumed to apply. 

5.11 FACILITY-SPECIFIC RADIONUCLIDE CONVERSIONS

The natural uranium fuel elements for the BGRR were reportedly fabricated by the BNL metallurgy 

group. The enriched uranium, aluminum clad elements which replaced the natural fuel were 

apparently not fabricated at BNL. These data are summarized in Table 5-6. 

Table 5-6. Facility-specific radionuclide conversions. 

Process description

Activity per unit mass (Bq/g uranium)

U-234 U-235 U-236 U-238

BNL metallurgy group, natural uranium fuel 

fabrication, late 1940s and possibly up to 1958

1.26E+04 5.59E+02 negligible 1.21E+04

5.12 WORKPLACE MONITORING DATA

If bioassay data are not adequate to evaluate an individual’s internal doses, dose reconstructors can 

use workplace monitoring data (NIOSH 2002). The following types of workplace data might be 

available for BNL: breathing zone air samples, general area air samples, and surface contamination 

surveys. However, these data are not likely to be in individual exposure records. Data on respirator 

use are not likely to be available. Quantitative fit testing information was not located. In the case of 

surface contamination data, site/process-specific resuspension factors are not likely to be available. 

During the operation of the HFBR, the airborne tritium contamination had been between 1 × 10-6

μCi/cm3 and 1 × 10-5 μCi/cm3 on the equipment level. Since the shutdown, the concentration has 

been between 1 × 10-7 μCi/cm3 and 5 × 10-7 μCi/cm3 on the equipment level and between 1 × 10-8

μCi/cm3 and 5 × 10-8 μCi/cm3 on the operations level. A reasonable estimate of the maximum 

exposure can be made by assuming that an individual spends 4 hours/day on the operations level and 

4 hours/day on the equipment level (BNL 1999a).

5.13 SOURCE TERM DATA

Without bioassay or air sample data, the last resort is determination of airborne concentrations using 

source term evaluations (NIOSH 2002, p. 19). Data on the amount of dispersible material available 

does not appear to be available for BNL.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 85 of 106

5.14 RADON

For dose reconstruction under EEOICPA, occupational radon exposure is exposure to radon 

emanating from sources other than those naturally occurring in the area. Dose reconstructors must 

subtract the natural background level of radon exposure from any measured values when assessing 

occupational exposure (NIOSH 2002, p. 32). BNL was not a processing or storage location for large 

quantities of 226Ra or 222Rn.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 86 of 106

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6.0

6.1 INTRODUCTION

OCCUPATIONAL EXTERNAL DOSIMETRY

Brookhaven National Laboratory (BNL) was established to provide facilities for scientific research 

which, because of size, complexity or mode of operation were beyond the means of most single 

universities. In providing nuclear reactors and various particle accelerators for its own staff and 

perhaps more importantly visiting scientists emphasis was placed on a multidisciplinary approach to 

scientific questions. That same mode of operation is still in practice today.

6.2 PURPOSE

The purpose of this section of the BNL site profile is to describe the external dosimetry systems and 

practices at BNL, and to assist in the evaluation of occupational external exposures from processes 

that occur at BNL. An objective of this section is to provide supporting technical data to evaluate, with 

assumptions favorable to the claimant, occupational external doses that can reasonably be 

associated with radiation exposures to both the monitored and unmonitored worker. 

6.3 SCOPE

This section presents historical and current practices in relation to the evaluation of external exposure 

data for monitored and unmonitored workers, and addresses the evaluation of worker exposure, 

missed dose, and the bias and uncertainty associated with the monitoring of external dose.

6.4 BNL DOSIMETRY PROGRAM

The Oak Ridge dosimeter was used and processed by BNL from start-up through 1984, at which time 

they switched to a vendor (Landauer, Lane 2003). The vendor services continued through 1995.

In December 1995, BNL was DOELAP accredited (Loesch 1995) and BNL started its own program 

using the Harshaw 8814 and 8806 TLD dosimeters. At times, primarily in the accelerator facilities, 

these dosimeters were complemented with CR-39 as the neutron dosimeter (Sengupta 2000). 

6.4.1

The neutron doses were measured with NTA film from the very beginning through 1995 when it was 

replaced with TLD. In addition, CR-39 and Lexan are also used in neutron dosimetry. The CR-39 

and Lexan are analyzed by an outside vendor. Spallation products produced stars on the NTA film 

and were evaluated by counting the number of prongs on the stars that were produced by those 

spallation products. The total number of prongs were counted and related to an effective neutron 

dose with the result added to the total neutron dose. It was found that the dose from these spallation 

products was very small. A RBE and later a QF of ten was used from the very beginning of 

operations of the various accelerators and cyclotrons for these products (Cowan 1953).

Neutron Dosimetry

6.4.2

On occasion special dosimeters were worn to monitor non-routine work that resulted in “significantly 

non-uniform doses to various areas of the whole body” (Sengupta 2000). The special dosimeters 

were worn without the regular dosimeter. The highest dose measured by the special dosimeters was 

the dose of record. In cases of lost or destroyed dosimeters, results were derived from past results of 

similar work, co-worker results or instrument measurements and time spent in the radiation zone. All 

Special Dosimetry
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 92 of 106

radiations measured were recorded as WB dose or dose equivalent. The correct value for the RBE 

(later the QF) was determined to be conservative; since a value of 10 was always used (Patterson 

1964, Cowan 1963). 

6.4.3

In the cases of lost or destroyed dosimeters, results were derived from past results of similar work, coworker results or the product of instrument measurements and time spent in the radiation zone. 

Lost or Destroyed Dosimeters

6.5 MISSED DOSE

The missed doses at BNL are basically those resulting from dosimeter minimum detectable levels 

(MDLs), exchange frequencies, and the possible error in determining the true doses due to the 

interaction of high energy particles. In the later case it was determined that the values of the RBEs 

and later QFs were always conservative resulting in larger than true doses recorded (Cowan 1953, 

Cowan unknown). It also appears that all workers were issued dosimeters from startup through 1954. 

Missed doses for monitored employees at BNL are basically from dosimeter MDLs and exchange 

periods. In reviewing individual dose records, the exchange period and zeros for that individual can 

be determined. The review of individual records is necessary because exchange periods varied prior 

to 1955 (Cowan 1953). This is applicable for both Hp (0.07) and Hp (10). 

Table 6-1 lists the period of use, type of dosimeter, exchange period, MDL, and estimated annual 

missed dose. 

Table 6-1. Estimated annual missed photon, beta, and neutron dose (mrem). 

Period of use Dosimeter

MDL

(mrem)a 

Exchange

frequency

Maximum annual

missed dose (mrem)b 

Startup through 1954 Multi-element film + NTA 30 Weekly

Monthly

780

180

1955 through 1995 Multi-element film +NTA 30 Monthly 180

1996 to present Harshaw 8814 TLD+ CR-39 10 Monthly 60

Startup through 1995 NTA filmc <50 Weekly

Monthly

1,300

300

1996 to present CR-39c <20 Monthly 120

a. Estimated MDLs for each dosimeter in the workplace even though many doses were reported at less than the MDL.

b. Estimated annual missed dose calculated using MDL/2 from NIOSH (2002).

c. Processing done by RS Landauer.

6.5.1

Unmonitored workers are those workers who did not enter radiation zones, or areas, where work with 

radioactive materials was undertaken. These workers were not issued dosimeters if the probability of 

exceeding 10% of the allowable limit was small (Sengupta 2000). Therefore these workers may be 

assigned doses as given in Table 4-1 in Section 4 of this document. All radiation workers were 

monitored from the startup of operations (Cowan, 1953). 

Unmonitored Worker

Missed doses for unmonitored employees could be as much as 500 mrem or 10% of whatever 

standard was in effect at the time of employment. Workers who did not enter radiation zones or areas 

where work with radioactive materials was undertaken did not receive dosimeters if the probability of 

exceeding 10% of the allowable limit was small (Sengupta 2000). 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 93 of 106

6.6 RADIATION ENERGIES AND PERCENTAGES AT SELECTED BNL FACILITIES 

Section 2, describes the different types of facilities at BNL including reactors, accelerators and various 

support facilities and target areas.

Tables 6-2 and 6-2a list the facilities and related data.

Table 6-2. Selection of beta (electron) and photon energies and percentages. 

Process/building Description

Operations Radiation 

type

Energy

Begin End keV Percentage

BGRR Reactor 1950 1969 Beta

Photon

>15

30-250

>250

100%

25%

75%

HFBR/750 Reactor 1965 1999

BMRR/491 Reactor 1959 2000

MRC490/490A Radio-isotopes 1958 Present

RTF/490 Accelerators Electron

Photon

>15

>250

100%

100%

BLIP/931 Accelerator 1973 Present Electron

Photon

>15

30-250

>250

100%

a 

a

PET/490 Accelerator 1977 Present Electron

Photon

>15

>250

100%

100%

Cosmotron/same Cyclotron 1952 1966 Photon 30-250

>250

a

RHIC/same Accelerator 1999 Present a

AGS/same Synchrotron 1960 Present Photon 30-250

> 250

a

AGS experimental area/912 Accelerator a 

AGS ring/913

AGS booster/942 1992 Present

200 MeV Linear Accelerator/

930

Accelerator 1972 Present

Tandem Van de Graaff/901 Accelerator 1970 Present

Tandem-to-Booster Beam 

Line/ Heavy Ion Beam tunnel

Accelerator Unknown

Accelerator Test Facility/ 820 Accelerator Electron

Photon

>15

unknown a

Nat’l. Synchrotron Light 

Source/725 

Electron storage 1982 Present Electron

Photon

>15

30-250

100%

100%

Chemistry Dept./490A Accelerator Electron

Photon

>15

unknown

100%

a

a. Unknown, dependent on target material and isotope produced. Most likely 100% >250 

Table 6-2a. Beta and photon energies and percentages (support facilities).

a

Process/building Description

Operations Radiation 

type

Energy

Begin End keV Percentage

NSRL NASA Space Lab 2003 Unknown Unknown Unknown

Tritium evaporator/802B Tritium processing 1995 Beta <15 avg. 100%

WCF/811 Waste processing Beta

Photon

>15

30-250

>250

100%

25%

75%

TPL/801 Hot Laboratory

HWMF/444 Incinerator

Chemistry Dept/555 & 490A Chemistry R&D

LEAF/555 Laser electron 

accelerator

Biology/463 Biological R&D

Oceanographic Sciences/318 Oceanographic 

R&D

Rad. Sciences/703,703W HP & Env. Support

Nuclear Waste Mgt./830 Waste Handling
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 94 of 106

a. The isotopes involved ranged from H-3 to Au-198 resulting in beta, electron, and photon energies ranging from a few 10s to 

several 1000s of keVs with wide ranging percentages.

6.7 NEUTRON RADIATIONS AND PERCENTAGES

Table 6.3 lists facilities with neutron radiations. They are reactors, accelerators, general laboratories, 

and medical facilities. The reactors are no longer in operation, the last being shut down in 2000.

Table 6-3. Facilities, neutron energies, percentages, and correction factors. 

Facility Source

Neutron energy

(MeV)

Default dose % and 

correction factor

Reactors- BGRR, HFBR, BMRR Reactors 0.1 to 2.0, Wr=20 100% correction factor = 1.91

Accelerators-Cosmotron, AGS, 

RHIC, NSLS, Medical

Particle reactions 0.1 to 2, Wr= 20

2 to 14+

100% correction factor = 2.0b

100% correction factor = 1.32

Sources- Medical, Calibration, 

Chemistry

Sources-AmBe, 

PuBe, Cf-252

0.1 to 2.0(0.57)a

2 to 14 (.43)a 

100% Cf-252, correction 

factor=1.91

100% Am or Pu Be or accelerator 

> 2 MeV, correction factor =1.32

Avg. correction if both Cf-252 or 

either AmBe or PuBE sources are 

present = 1.66

a. Weighted average dose from alpha-neutron and SF (spontaneous fission) neutrons (Y-12 Section 6 TBD, ORAU-TKBS0014-6,10/11/05).

b. Correction factor for those exposures resulting from accelerator operations was always 10 be it either RBE or QF. 

All potential neutron exposures of >100 mrem annually used NTA film through 1995. Starting in 1996 

the Harshaw 8806 TLD along with CR-39 was used at the accelerators (Lane 2003, Sengupta 2000). 

Recorded neutron doses include both thermal and fast neutrons generated by the various 

accelerators with energies >100 MeV (Sengupta 2000). The actual “dose of record” always included 

a QF of 10 starting from the first day of operation. 

6.8 RECORDED DOSE PRACTICES

Recorded dose practices at BNL are given in Table 6-4 and include those provided by both the site 

and its vendor and any special dosimeter results (Sengupta 2000). Special dosimeters were worn if 

needed due to abnormal conditions, such as highly directional beams. If more than one dosimeter 

was required the highest value was recorded. The regular dosimeter was not worn on these 

occasions. 

Table 6-4. Recorded dose practices. 

1949 through 1984 Oak Ridge film 

dosimeter + NTA

Beta=Open window, mrem

Photon (P), mR

Fast Neutron (FN), mrem

Skin= OW+P

WB= P+ FN+ tritium 

1985 through 1995 Landauer + NTA film Beta or nonpenetrating, mrem

Photon (P), mrem 

Fast Neutron (FN), mrem 

Skin= NP + P 

WB= NP + P + FN + tritium 

1996 to present – Harshaw 8814 & Harshaw 

8806a TLD + CR-39b for neutrons 

Beta or nonpenetrating, mrem

Photon (P), mrem

Fast Neutron (FN), mrem 

Skin= NP + WB

WB= P + NP + FN + tritium 

a. Issued only when higher energy neutrons might be present.

b. Attached to either dosimeter when neutrons might be present.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 95 of 106

6.9 INTERPRETATION OF REPORTED DOSES.

Early personnel doses were reported in units of mrep for both penetrating (photon) and 

nonpenetrating (beta) if beta or low energy electrons were present. High energy beta or electrons 

were recorded as penetrating dose. The recorded total doses include the results from any special 

dosimeter worn for that exchange period. If it is necessary to obtain organ doses, dose reconstructors

should use External Dose Reconstruction Implementation Guidelines (NIOSH 2002). 

Table 6-5 lists the interpretation of the reported data by time period. Reported doses have been 

corrected for background using site controls. The controls are dosimeters kept on site in locations 

used for the storage of personnel dosimeter. All issued dosimeters were stored on storage racks 

located at various positions through-out the site. Dosimeters were not to be taken home at the end of 

the shift, a practice beginning at the start-up of the site and continues at present (Sengupta 2000).

Table 6-5. Interpretation of reported data. 

Period

Reported 

quantity Description

Interpretation 

of zeros

Interpretation 

of blanks

Rollup of 

individual and 

annual data Monitored/unmonitored

Startup to 

1985

Skin= mrad

WB=mR

Neutrons=mrem

mrep (about startup)

mR, mrad and mrem 

used interchangeably

MDL/2 times 

number of 

zeros.

If no dosimeter 

for that period 

treat as 

unmonitored.

If special 

dosimeters 

were used, 

include results.

All personnel expected to be 

exposed to >100 mrem in 

an exchange period were 

monitored.

1986 to 

present

Skin=mrem

WB= mrem

Neutrons=mrem

mrem used for all. MDL/2 times 

number of 

zeros.

If no dosimeter 

for that period 

treat as 

unmonitored.

If special 

dosimeters 

were used, 

include results.

Only those>10% of current 

standards were monitored. 

Those entering controlled 

areas were issued visitors 

dosimeter.

6.10 ADJUSTMENTS TO RECORDED DOSE

Since most but not all penetrating photons are above 30 keV, it is suggested that an adjustment is 

necessary to account for the contribution to Hp(10) from the lower energy photons. It is estimated 

that a correction equal to 10% of the <250-keV values be made. Corrections, where required, can be 

made to the WB dose using the product of 10% times the percentage <250 keV. This adjustment 

equals 2.5% for those facilities having photons in the 30 to 250 keV range except the NSLS which is 

10% since most all of it’s photons are estimated to be below 250 keV (measurements indicate a peak 

at 75 keV (Preisig 1997). These adjustments also increase Hp(0.07) because that dose was always 

the sum of the OW and WB doses. Table 6-6 lists the corrections.

Table 6-6. Adjustments to recorded dose. 

Period Dosimeter Facility Adjustment to reported dose

Startup to 1985 Oak Ridge Reactors Multiply reported dose by 1.025 to 

estimate Hp(10)

1958 to present Oak Ridge, Landauer

and Site

MRC 490/490A Multiply reported dose by 1.025 to 

estimate Hp(10)

1982 to present Oak Ridge, Landauer,

Site

NSLS Multiply reported dose by 1.10 to 

estimate Hp(10)

6.11 BIAS AND UNCERTAINTY

Bias and uncertainty values were not found in any site data. Values given in Table 6-7 are those 

found in the Hanford TBD (ORAU-2004) and should be similar to those at BNL since both used the 

multi-element film dosimeter and the TLD dosimeter. 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 96 of 106

The Landauer dosimeter values were taken from the Atomics International (AI) TBD* (ORAU -2006). 

The NTA film values were also taken from the AI TBD because the adjusted neutron energy values 

are similar and are based on Y-12 data (ORAU-2005).

* The Atomics International TBD was originally the ETEC TBD
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 97 of 106

Table 6-7. Bias and uncertainty. 

Site specific 

dosimetry system

Bias magnitude and range Uncertainty factors

Overall biasb Range in bias Systematicc Randomd

Multi element

Start-up thru 1995a 1.02 0.86-10 1.1 1.4

Site TLD 1.01 0.95-1.05 1.05 1.2

NTA film

0.1 to 2 MeV

2.0 to 14 MeV

CR-39

1.35

1.35

Unknown

0.5 –1.5

0.65-1.35

Unknown

1.5

1.35

Unknown

a. These values agree with Landauer since NVLAP certification (Yoder 2005).

b. Based on most likely distribution of energy levels and geometries. Divide recorded dose by 

table’s bias value to determine deep dose.

c. Systematic uncertainty due to lack of knowledge of actual distributions of energies and 

geometries.

d. Random uncertainty due to variations among workers in energy levels and geometry.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 98 of 106

REFERENCES

Brookhaven National Laboratory Radiological Control Manual, Rev. 2, Nov. 20, 1995, Data Capture 

Document Discovery, 3/2/2006. 

Cowan, Frederick P., and J.S. Handloser, Health Physics Program for the Brookhaven Cosmotron, 

BNL-264 CT-43, Nov. 1953, Data Capture Discovery Document. 

Cowan, Frederick P., A Preliminary Report On Health Physics Problems At The Brookhaven 

Alternating Gradient Synchrotron, unknown, Data Capture Discovery Document, 3/3/2006.

Cowan, Frederick, Leigh F. Phillips, and Robert J. King, Some Dosimetry Problems of the Alternating 

Gradient Synchrotron (AGS), BNL 7151, July 1, 1963, Data Capture Discovery Document, 

3/3/2006. 

Cowan, Frederick P., Personnel Dosimetry of Very-High Energy Radiations, Symposium On 

Personnel Dosimetry Techniques For External Radiation, Madrid, 1-5 April 1963, Data Capture 

Discovery Document, 3/3/2006. 

Cowan, Frederick P., Health Physics Program For The Brookhaven National Laboratory 

Synchrotrons, Unknown, Data Capture Document Discovery, 3/3/2006. 

Handloser, John S., Stray Radiation Dosimetry Around Pulsed Accelerators, BNL 3519, Unknown., 

Data Capture Discovery Document, 3/3/2006. 

Holeman, G.R., Personnel Monitoring External Dosimetry Quality Assurance Plan, Procedure No. PMQAP-001, Nov. 20,1996, Data Capture Document Discovery, 3/2/2006. 

Lane, Georgette, BNL e-mail, External Dosimetry and Tritium Dose at Brookhaven, Dec. 22, 2003, 

Data Capture Discovery Document 030004414.

Loesch, Robert M., Laboratory Accreditation of Brookhaven National Laboratory (BNL) Personnel 

Dosimetry, Dec. 12, 1995, Letter to Carson L. Nealy, DOE/BAO, Data Capture Discovery 

Document, 3/3/2006.

NIOSH ( National Institute for Occupational Safety and Health),2002, External Dose Reconstruction 

Implementation Guidelines, OCAS-IG-001, Rev.1, Office of Compensation Analysis and 

Support, Cincinnati, Ohio. 

ORAU (Oak Ridge Associated Universities), 2006, Technical Basis Document for the Energy 

Technology Center-External Dosimetry, ORAUT-TKBS-0038-6, Oak Ridge Tennessee.

ORAU (Oak Ridge Associated Universities), 2004, Technical basis Document for the Hanford SiteExternal Dosimetry, ORAUT-TKBS-0006-6, Oak Ridge, Tennessee.

ORAU ( Oak Ridge Associated Universities), 2005, Technical Basis Document for the Y-12 Security 

Complex-Occupational External Dosimetry, ORAUT-TKBS-0014-6, Oak Ridge, Tennessee.

Patterson, H. Wade and Ralph H. Thomas, Editors, The History of Accelerator Radiological 

Protection, Nuclear Technology Publishing, ISBN 1 870965 31 0, 1994. 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 99 of 106

Preisig, Joseph R., 1995 Alternating Gradient Synchrotron (AGS) Bonner Multisphere Spectrometry 

and Carbon-11 Detector Results, Letter to Edward Lessard, AGS Department, July 14, 1995, 

Data Capture Document Discovery, 3/29/2006.

Preisig, Joseph R., 1997 Bonner Multisphere Spectrometer Measurements at the NSLS (April/May, 

1997); HP-51-20-97, Data Capture Document Discovery, 3/29/2006.

Reciniello, Richard N., Un-Monitored Workers, e-mail to L Faust, Brookhaven National Laboratory, 

4/4/2006. 

Sengupta, S., External Dosimetry Technical Basis Document, March 31, 2000, PM-PD-003, Revision 

1, Data Capture Discovery Document, 2/28/2006.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 100 of 106

GLOSSARY

absorbed dose

Amount of energy in rads or grays deposited in a substance by ionizing radiation per unit mass 

of the substance. See dose. 

absorption type 

Categories for materials according to their speed of absorption in the body, which replaced the 

earlier inhalation clearance classes. Defined by the International Commission on Radiological 

Protection, the absorption types are F for fast absorption (formerly inhalation class D), M for 

moderate absorption (formerly inhalation class W), and S for slow absorption (formerly 

inhalation class Y). 

accelerator

See particle accelerator. 

accreditation

Recognition that a dosimeter system has passed the applicable performance criteria of a 

program such as the U.S. Department of Energy Laboratory Accreditation Program or the 

former National Voluntary Laboratory Accreditation Program.

accuracy

Extent to which a given value agrees with the standard value for that measurement. If a series 

of measurements has small systematic (nonrandom) errors, they are said to have high 

accuracy. The standard deviation of a series of values is a measure of the accuracy of those 

values.

Atomic Energy Commission (AEC)

An agency established by the U.S. Government for oversight of nuclear weapons and power 

production; a predecessor to the U.S. Department of Energy.

backscatter

Reflection or refraction of radiation at angles over 90 degrees from its original direction.

beta particle (β) 

See beta radiation. 

beta radiation 

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron. Most of the direct fission products are (negative) beta emitters. 

Exposure to large amounts of beta radiation from external sources can cause skin burns 

(erythema), and beta emitters can be harmful inside the body. Thin sheets of metal or plastic 

can stop beta particles.

cladding

The outer layer of metal that encases a reactor fuel element or fissile material of the pit of a 

nuclear weapon, often made with aluminum or zirconium. In a reactor, cladding promotes the 

transfer of heat from the fuel to the coolant, and it builds up fission and activation products 

over time from the fission of the fuel.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 101 of 106

core

Central region of a nuclear reactor where fission of the fuel takes place. 

Cosmotron (Synchrotron) 

Accelerators in which charged particles (electrons, protons, ions, etc.) are accelerated in a 

circular path achieving very high energies.

criticality

State of a radioactive mass (e.g. the core of a nuclear reactor) when the fission reaction 

becomes self-sustaining. Nuclear reactors go critical when started.

cyclotron

Particle accelerator capable of large beam currents where the beam is injected in the center of 

a circular magnet. A fixed radio frequency signal applied to two D-shaped electrodes 

accelerates the beam as it passes from one electrode to the other as the potential alternates. 

The radius of the beam increases as the energy increases.

curie (Ci) 

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

deep dose equivalent [Hp(10)]

Dose equivalent in units of rem or sievert for a 1-centimeter depth in tissue (1,000 milligrams 

per square centimeter). See dose. 

dose 

In general, the effects of ionizing radiation in terms of the specific amount of energy absorbed 

per unit of mass. Effective and equivalent doses are in units of rem or sievert; other types of 

dose are in units of roentgens, rads, reps, or grays. 

dose equivalent (H)

Product of absorbed dose in units of rem or sievert in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose. 

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual.

dosimetry

Measurement and calculation of internal and external radiation doses.

dosimetry system

System for assessment of received radiation dose. This includes the fabrication, assignment, 

and processing of external dosimeters, and/or the collection and analysis of bioassay samples, 

and the interpretation and documentation of the results. 

electron

Basic atomic particle with negative charge and a mass 1/1,837 that of a proton. Electrons 

surround the positively charged nucleus of the atom.
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 102 of 106

exchange period (frequency)

Period (weekly, biweekly, monthly, quarterly, etc.) for routine exchange of dosimeters.

exposure

(1) In general, the act of being exposed to ionizing radiation. (2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens.

favorable to the claimant

In relation to dose reconstruction for probability of causation analysis, having the property of 

ensuring that there is no underestimation of potential dose, which often means the assumption 

of a value that indicates a higher dose than is likely to have actually occurred. See probability 

of causation. 

film

Radiation-sensitive photographic film in a light-tight wrapping.

film dosimeter

Package of film for measurement of ionizing radiation exposure for personnel monitoring 

purposes. A film dosimeter can contain two or three films of different sensitivities, and it can 

contain one or more filters that shield parts of the film from certain types of radiation. When 

developed, the film has an image caused by radiation measurable with an optical 

densitometer.

filter

Material used in a dosimeter to adjust radiation response to provide an improved tissue 

equivalent or dose response.

fission

Splitting of the nucleus of an atom (usually of a heavy element) into at least two other nuclei 

and the release of a relatively large amount of energy. This transformation usually releases 

two or three neutrons. 

fission product

Radionuclides produced by fission or by the subsequent radioactive decay of radionuclides. 

gamma radiation 

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma rays are very 

penetrating, but dense materials such as lead or uranium or thick structures can stop them. 

Gamma photons are identical to X-ray photons of high energy; the difference is that X-rays do 

not originate in the nucleus. 

gamma ray

See gamma radiation. 

gray (Gy)

International System unit of absorbed radiation dose, which is the amount of energy from any 

type of ionizing radiation deposited in any medium; 1 Gy equals 1 joule per kilogram or 

100 rads. 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 103 of 106

in vitro

Of or relating to a process that takes place under artificial conditions or outside a living 

organism (e.g., in the laboratory). From Latin meaning in glass. 

in vivo

Of or relating to a process that takes place in a living organism. From Latin meaning in life. 

ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons. See 

alpha radiation, beta radiation, gamma radiation, photon radiation, and X-ray radiation. 

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties but often have different physical 

properties.

kerma

Measure in units of absorbed dose (usually grays but sometimes rads) of the energy released 

by radiation from a given amount of a substance. Kerma is the sum of the initial kinetic 

energies of all the charged ionizing particles liberated by uncharged ionizing particles 

(neutrons and photons) per unit mass of a specified material. Free-in-air kerma refers to the 

amount of radiation at a location before adjustment for any external shielding from structures 

or terrain. The word derives from kinetic energy released per unit mass. 

lumbar spine

The vertebrae of the lower back.

natural uranium

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

weight. The specific activity of this mixture is 2.6 × 107 becquerel per kilogram (0.7 picocuries 

per gram).

neutron

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen.

nonpenetrating dose (NP, NPEN) 

Dose from beta and lower energy photon radiation, often determined from the open-window 

dose minus the shielded window dose. See dose. 

nuclide 

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide.
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open window (OW)

Area of a film dosimeter that has little to no radiation shielding (e.g., only a holder and visible 

light protection). See film dosimeter. 

particle accelerator 

Device that accelerates (imparts energy to) ions and magnetic or electrostatic fields for 

focusing and redirecting ion beams. The main purposes of accelerators are the investigation 

of high-energy particle behavior and synthetic isotopes.

penetrating dose (PEN) 

Dose from higher energy photon (gamma and X-ray) radiation that penetrates the outer layers 

of the skin. See dose. 

personal dose equivalent [Hp(d)] 

Dose equivalent in units of rem or sievert in soft tissue below a specified point on the body at 

an appropriate depth d. The depths selected for personal dosimetry are 0.07 millimeters 

(7 milligrams per square centimeter) and 10 millimeters (1,000 milligrams per square 

centimeter), respectively, for the skin (shallow) and whole-body (deep) doses. These are 

noted as Hp(0.07) and Hp(10), respectively. The International Commission on Radiological 

Measurement and Units recommended Hp(d) in 1993 as dose quantity for radiological 

protection. 

photon

Basic unit of electromagnetic radiation. Photons are massless “packages” of light energy that 

range from low-energy microwave photons to high-energy gamma rays. Photons have 

energies between 10 and 100 kiloelectron-volts. See photon radiation. 

photon radiation 

Electromagnetic radiation of light energy (photons) from microwaves to gamma rays. Gamma 

rays and X-rays are examples of ionizing photon radiation, which have enough energy to 

penetrate matter, including the body, and deposit energy in that matter. 

PM

A procedure detailing specific actions or directions and usually limited to one service or 

activity.

probability of causation

For purposes of dose reconstruction for the Energy Employees Occupational Illness 

Compensation Act, the percent likelihood that a worker incurred a particular cancer from 

occupational exposure to radiation.

rad

Traditional unit for expressing absorbed radiation dose, which is the amount of energy from 

any type of ionizing radiation deposited in any medium. A dose of 1 rad is equivalent to the 

absorption of 100 ergs per gram (0.01 joules per kilogram) of absorbing tissue. The rad has 

been replaced by the gray in the International System of Units (100 rads = 1 gray). The word 

derives from radiation absorbed dose. 
Document No. ORAUT-TKBS-0048 Revision No. 00 Effective Date: 08/30/2006 Page 105 of 106

radiation

Subatomic particles and electromagnetic rays (photons) that travel from one point to another, 

some of which can pass through or partly through solid materials including the human body. 

See ionizing radiation. 

radioactivity

Disintegration of certain elements (e.g., radium, actinium, uranium, and thorium) accompanied 

by the emission of alpha, beta, gamma, and/or neutron radiation from unstable nuclei. See 

radionuclide. 

radiograph 

Photographic image produced on film by gamma rays or X-rays. Some of the rays (photons) 

can pass through parts of an item, while more opaque parts partially or completely absorb 

them and thus cast a shadow on the film. See radiology. 

radiology 

Science of using X-rays to produce images on photographic film, especially for medical 

purposes. 

radionuclide

Radioactive nuclide. See radioactive and nuclide. 

rem 

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The average American receives 360 millirem a year from background radiation. The 

sievert is the International System unit; 1 rem equals 0.01 sievert. The word derives from 

roentgen equivalent in man; rem is also the plural.

rep 

Historical quantity of radiation (usually other than X-ray or gamma radiation) originally defined 

as 83 ergs absorbed per gram in the body and redefined in the 1940s or early 1950s as the 

amount that would liberate the same amount of energy (93 ergs per gram) as 1 roentgen of Xor gamma rays. Replaced by the gray in the International System of Units; 1 rep is 

approximately equal to 8.38 milligray. The word derives from roentgen equivalent physical. 

roentgen (R) 

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

or negative charge equal to 2.58 × 10-4 coulombs per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0°C and standard atmospheric pressure. An 

exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft tissue for 

higher energy photons (generally greater than 100 kiloelectron-volts). 

sievert (Sv) 

International System unit for dose equivalent, which indicates the biological damage caused 

by radiation. The unit is the radiation value in gray (equal to 1 joule per kilogram) multiplied by 

a weighting factor for the type of radiation and a weighting factor for the tissue; 1 Sv equals 

100 rem.

skin dose

Absorbed dose at a tissue depth of 7 mg per square centimeter.
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thermoluminescent dosimeter (TLD) 

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated by radiation, release the stored energy as light. The measurement of this 

light provides a measurement of absorbed dose. 

tissue equivalent

Substance with response to radiation equivalent to tissue. A tissue-equivalent response is an 

important consideration in the design and fabrication of radiation measuring instruments and 

dosimeters.

type

See absorption type. 

whole-body dose

Dose to the entire body excluding the contents of the gastrointestinal tract, urinary bladder, 

and gall bladder and commonly defined as the absorbed dose at a tissue depth of 

10 millimeters (1,000 milligrams per square centimeter). See dose. 

X-ray 

(1) See X-ray radiation. (2) See radiograph. 

X-ray radiation 

Penetrating electromagnetic radiation (photons) of short wavelength (0.001 to 10 nanometers) 

and energy less than 250 kiloelectron-volts. X-rays usually come from excitation of the 

electron field around certain nuclei. Once formed, there is no difference between X-rays and 

gamma rays, but gamma photons originate inside the nucleus of an atom. 

zirconium

Metallic element with atomic number 40. Zirconium is highly resistant to corrosion, and it is 

alloyed with aluminum to make cladding for nuclear fuel and sometimes in small amounts with 

the fuel itself.